核动力工程

Present Situation and Prospect of Radioactive Waste Liquid Treatment Technology

Sun Shouhua, Ran Mingdong, Lin Li, Liu Wenlei, Li Zhenchen, Li Wenyu

2019, 40(6): 1-6. doi: 10.13832/j.jnpe.2019.06.0001

Abstract(1246) PDF(726)

Abstract:
The effective disposal of radioactive waste liquid is the precondition for the rapid development of nuclear industry all over the world, and the current situation and development direction of its key technologies are the focus of attention of the nuclear industry in China. This paper introduces several traditional methods of radioactive waste liquid treatment and the emerging new technology options, summarizes the principles, advantages and disadvantages of various methods, and discusses the research direction and development trend of radioactive waste liquid treatment technology in the future.

Prediction of Annular Flow in Vertical Round Tube Based on Two-Phase CFD Approach

Xu Haisong, Wang Ji, Xiong Jinbiao, Lu Chuan

2019, 40(6): 7-12. doi: 10.13832/j.jnpe.2019.06.0007

Abstract(245)

Abstract:
Based on the Euler-Lagrangian method and the liquid film model, the phase interaction phenomena, including the interaction between the liquid drop and continuous gas, liquid drop deposition and entrainment on the wall and the liquid film boiling and evaporation, are modelled for the annular flow in a round tube. The prediction accuracy of the Euler-Lagrangian method for the annular flow is evaluated by comparing the results of the annular flow experiment of KTH. The comparison results show that the mass flow rate of the liquid film obtained with numerical simulation agrees well with the experimental results. The results also verify that the Euler-Lagrangian method, accompanied with the applied flow and heat transfer model, can reasonably simulate the annular flow.

Research on Neutronics Characteristics of Small Modular Super-Safe Gas-Cooled Reactor

Zhang Chenglong, Du Shuhong, Liu Guoming, HuoXiaodong, Yang Haifeng, Shao Zeng

2019, 40(6): 13-19. doi: 10.13832/j.jnpe.2019.06.00013

Abstract(276)

Abstract:
To analyze the neutronics characteristics of the small modular super-safe gas-cooled reactor, this paper investigates how some physical parameters, such as TRISO particle packing fraction and fuel enrichment, influence the burnup characteristics, based on the hexagonal prism fuel assembly model, using Monte Carlo procedure and ORIGEN procedure. The results show that the lifetime increases with the fuel enrichment and pitch-to-diameter ratio, and is influenced by the diameter of fuel rod, the packing fraction and size of TRISO particles to a certain extent, while the effect of the TRISO particle coating layer thickness on the lifetime can be neglected. Based on the above results, a preliminary core loading scheme for the reactor is proposed, which meets the requirement of a 20-year lifetime without refueling.

Study on Water Film Breakdown Phenomenon at Vertical Corrugated Plate Corner

Wang Bo, Tian Ruifeng

2019, 40(6): 20-23. doi: 10.13832/j.jnpe.2019.06.0020

Abstract(160)

Abstract:
Through the liquid film rupture model and the reduce of Navier-Stokes (N-S) equation and continuity equation for the 2-dimensional boundary layer in the curvilinear coordinate system, with non-dimension definition, the theory method for calculating the critical airflow velocity of water film breakdown is obtained; and, applying the Planar Laser Induced Fluorescence (PLIF) method and image processing method, the relationship between the water film thickness and the critical airflow velocity of water film breakdown is investigated through the water film breakdown experiment. Both experimental and theoretical calculation results show that the critical airflow velocity of water film breakdown is negatively correlated with the water film thickness. With the same water film thickness, the larger the critical breakdown structure parameter of corrugated plate dryer is, the greater the critical airflow velocity of water film breakdown is, which is beneficial to the improvement of the steam-water separation effect.

Experimental Study on Effect of Containment Debris on Pressure Drops of PWR Fuel Assembly after LOCA

Wang Da, Niu Fenglei, Liang Ruixian, Zhuo Weiqian, LiuJun

2019, 40(6): 24-29. doi: 10.13832/j.jnpe.2019.06.0024

Abstract(264)

Abstract:
To study the impurity accumulation phenomenon and corresponding impact on head loss through fuel assembly after loss of coolant accident in Pressurized Water Reactors (PWRs), the test platform is established and the distribution and blockage of debris in the assembly in the worst condition were analyzed, and the effect of the debris in the containment after LOCA on the pressure drops across a fuel assembly is quantatively evaluated. All test results indicated that almost all debris were captured on the lower half of the fuel assembly, especially on bottom nozzle. The head loss resulted from debris added with concurrent addition method was higher than that with sequential addition method with the same debris amount; The chemical precipitate could compact the debris bed, resulting in a great effect on head loss; Enough coolant could be provided into the core for cooling even in worst condition.

Analysis of Fluctuation Characteristics of Power Range Detector for CPR1000

He Yang, Zhang Haizhou, Cao Yunlong, Wang Zongkui

2019, 40(6): 30-34. doi: 10.13832/j.jnpe.2019.06.0030

Abstract(265)

Abstract:
Based on engineering practices, this paper summarized the fluctuation characteristics of power range detector (PRC) through of the data investigation and analysis, and proposed some laws of fluctuation of power range. Suggestions on how to dig the electric power margin is provided. At the same time, it proposed potential research directions for some problems in the study of the power range fluctuation.

Study and Design of Two-Phase Flow Experiment Loop in a Rectangular Mini-Channel

Zhang Jie, Chen Hongli, Zhang Xue

2019, 40(6): 35-39. doi: 10.13832/j.jnpe.2019.06.0035

Abstract(243)

Abstract:
In order to study the flow and heat transfer characteristics of narrow rectangular coolant channels in plate-fuel reactors, the research status and development trend of the heat transfer in mini-channels around the world were reviewed. Based on the related parameters and experiment requirements of NP engineering reactors, a set of experimental loops for thermal hydraulic two-phase flow was established to study and analyze the heat transfer characteristics in plate-fuel cooling channels. According to real reactor operational requirements, a test section consistent with the actual coolant channel was designed. The flow direction is vertically downward, and the double-sided heating method is adopted, and the flow passage gap size is 2.3 mm. The channel width is 67 mm (62 mm-heated width), and the runner length is 1000 mm (750 mm-heated length). It was verified experimentally. The results show that the apparatus is correct and feasible.

Investigation of Uncertainty Quantification Method of Thermodynamic Models Based on EM Algorithm

Li Dong, Zhao Boyang, Jin Di

2019, 40(6): 40-45. doi: 10.13832/j.jnpe.2019.06.0040

Abstract(247)

Abstract:
The widely used BEPU method is to quantify the important uncertain source (input uncertainty) associated with the power plant and models, and then propagate the uncertainty to obtain the double 95% uncertainty bands of the safety related parameters (such as the peak cladding temperature, PCT). However, as the important sources of uncertainty, some physical model parameters in the best estimation (BE) codes are often difficult to be directly measured in the experiments. The quantification process is mainly dependent on the expert judgment which lacks objectivity. In this paper, based on the Expectation Maximization (EM) algorithm and the response parameters measured directly from the separate effect test, the probability density distribution of the internal model parameters can be derived by solving the inverse problem. Then, the application of the method is performed by taking reflood phenomena as an example. Uncertainties of code internal boiling heat transfer and interfacial friction models are obtained.

Development and Application of Advanced Reactivity Measurement System

Wang Lihua, Yang Qingxiang, Du Bing, Yang Bo, Qin Yulong, Shi Jianfeng

2019, 40(6): 46-49. doi: 10.13832/j.jnpe.2019.06.0046

Abstract(201)

Abstract:
Comparing with traditional reactor control rod worth measurement methods, the rapid dynamic rod worth measurement method requires a kind of reactivity measurement equipment with higher accuracy and performance to obtain and deal with the current signal from ex-core detectors, and the spatial effects during the test should be corrected by additional core neutronics calculation, therefore, in this research the Advanced Reactivity Measurement System(SMART) is developed including the advanced physics test computer(APTC) and the dynamic rod worth measurement code (LIGHT), which can wholly support physics startup tests including the dynamic rod worth measurement test. In addition, a series of verification tests are performed for SMART, and it can be indicated that SMART is capable to complete all the physics startup tests, and the accuracy and performance can satisfy physics startup tests including the dynamic rod worth measurement test. Engineering application capability of SMART is also verified by successful application in 300MW nuclear power plants.

Analytical Study on Safety of HFETR Cobalt Control Rods

Liu Shuiqing, Kang Changhu, Yang Bin, Tan Jinghua, Zou Peng, Liu Hongqian, Song Yuge, Song Jiyang

2019, 40(6): 50-53. doi: 10.13832/j.jnpe.2019.06.0050

Abstract(250)

Abstract:
The cobalt control rods of High Flux Engineering Test Reactor (HFETR) have been in service for over 20 years. In this paper, the burnup of 59Co in HFETR and its effect on the worth of cobalt control rods are calculated and analyzed. The results show that the average burnup and maximum burnup of 9#~14# shim-rod is 4.02% and 5.45%, respectively, and that of 4# and 7# shim-rod is 6.45% and 10.38%, respectively. Considering the effect of 59Co burnup, the worth of 9#~14# shim-rod is almost unchanged, and the worth of 4# and 7# shim-rod decreases 0.15 βeff (for HFETR, 1βeff=0.0071). The sub-criticality decreases 0.16βeff because of 59Co burnup, but the position of the shut-down rod is almost unchanged. So, the effect of 59Co burnup is small, with no effect for the safety of HFETR operation.

Numerical Study on Mixing Vane in Fuel Rod Bundle Channel of Nuclear Reactors

Wang Ye, Sun Lanxin, Xu Changzhe, Hu Nan, Zheng Xu, Zhang Meng, Li Xiaochang

2019, 40(6): 54-58. doi: 10.13832/j.jnpe.2019.06.0054

Abstract(255)

Abstract:
Mixing vanes installed in the spacer grids are significant structures of fuel assemblies in nuclear reactors, and the behavior of which has vital effect on the thermal-hydraulic characteristics of the fuel rod bundle channel. Numerical analysis of the effect of the vane arrangement and the vane tip shape was carried out to investigate the 5×5 rod bundle with single-span spacer grid. The results showed that the pressure drop was almost unaffected by the arrangement of the mixing vane, while the flow pattern and heat transfer characteristics downstream of the spacer grids varied evidently as the arrangement changed. As the shape of the vane tip changed into an arc, the pressure drop was not significantly different from that of the typical one, but the heat transfer characteristics was obviously improved.

Non-Destructive Evaluation Technology for Irradiation Surveillance of Reactor Pressure Vessels

Shangguan Bin, Li Chengliang

2019, 40(6): 59-63. doi: 10.13832/j.jnpe.2019.06.0059

Abstract(407)

Abstract:
Neutron irradiation embrittlement of reactor pressure vessel (RPV) is the most important issue in nuclear safety, which affects the safety and economy of and public confidence in nuclear power plants. This paper introduces the traditional RPV irradiation surveillance method, discusses its limitations, combs the research progress and existing problems of RPV irradiation embrittlement non-destructive evaluation(NDE) technology, creatively puts forward a unified model for predicting the mechanical properties of RPV steel under neutron irradiation based on experiment and theoretical research, and forms a new NDE technology based on electromagnetic properties, which has a good engineering application prospects after further improvement. At the same time, this paper points out that the experimental study of electro-magnetic properties of RPV steel will not only help us to understand and recognize the irradiation embrittlement behavior of RPV steel in long service life, but also help to reveal the mechanism and enrich the basic theory of irradiation embrittlement.

Criticality Safety Calculation and Analysis of Fresh Fuel Element Transport Containers for High Temperature Gas-Cooled Reactor

Li Yinghong, Huang Hao, Zhou Rongsheng, Yang Xiaodong

2019, 40(6): 64-71. doi: 10.13832/j.jnpe.2019.06.0064

Abstract(200) PDF(125)

Abstract:
In this paper, the HTR-10 fuel pebbles were described by MCNP5 program which is based on the Monte Carlo method. Eight different fuel pebble models were constructed, and their keff values and computation time have been compared to obtain an optimal fuel pebble model. Then, using this optimal fuel pebble model, a HTGR fresh fuel elements transport container model was described by MCNP5. And a series of criticality influencing factors are calculated and analyzed, such as thickness of the neutron absorbing plate, wall thickness of the aluminum alloy tube, thickness of the outer container, buffer layer materials, reflector materials, deformation, structural loss, water submergence and infinite container arrays. The results showed that the HTGR fresh fuel elements transport container always remains in the criticality safety state, no matter under normal transportation conditions or under accidental transportation conditions. And its criticality safety index (CSI) can be set to zero.

Simulation Study of Uranium Electrodeposition Behavior in Spent Fuel Dry Process

Zhang Meng, Wang Jingyang, Sun Lanxin, Mi Zongliang, Zhou Yu, Hou Hongguo, Gao Yang, Jiao Caishan

2019, 40(6): 72-76. doi: 10.13832/j.jnpe.2019.06.0072

Abstract(269)

Abstract:
Based on multi-physics coupling software COMSOL, a three-dimensional numerical model of uranium chloride electrodeposition behavior, in which the electrolytic cell serves as anode and the double graphite rod serve as cathode, was established in LiCl-KCl molten salt system. With the research in the movement of uranium ions and the real-time calculation of the cathode geometry, the change of the thickness of uranium electrodeposition with time is worked out. The relationship between the uranium deposition behavior and the position of the cathode surface, molten uranium ion concentration, reaction temperature, average current density is obtained. In this research, the simulation results are compared with the experimental data of uranium chloride electrolysis. The calculations are well fitted to the experimental results, and the reliability of the uranium electrodeposition behavior simulation is proved. The simulation results can provide a design reference for the extraction of the uranium from the spent fuel by dry reprocessing.

Simulation Technology for Post Weld Heat Treatment of Steel Containers for Nuclear Power Plants

Chang Haijun, Hou Yanyan, Yang Xiaohua, Shen Mengling

2019, 40(6): 77-81. doi: 10.13832/j.jnpe.2019.06.0077

Abstract(218)

Abstract:
The post weld heat treatment process of the weld between the first ring and the second ring in the containment was simulated by SYSWELD software, and then the residual stress and deformation of the whole circle and segment of the containment was calculated and analyzed. Afterwards, the effectiveness of the simulation was proved by comparison with experimental residual stress results, and the rationality of the partial segmentation heat treatment method was determined by combining with the actual situation of the project. Therefore, the research provided an effective way for post weld heat treatment of the nuclear steel containment with large wall thickness and large volume.

Generating Seismic Time Histories Compatible with Multi-Damping Floor Response Spectra Using Hilbert-Huang Transform

Li Bo, Tang Jiawei, Dai Kaoshan, Shen Le

2019, 40(6): 82-88. doi: 10.13832/j.jnpe.2019.06.0082

Abstract(181)

Abstract:
The United States Nuclear Regulatory Commission (USNRC) specifies the requirements of seismic time histories for the seismic design and analysis of nuclear power plants, in which the seismic time histories are either compatible with single-damping target response spectrum or with multi-damping target response spectra. In the seismic design and analysis of nuclear equipment and components, the seismic time history compatible with the multi-damping floor response spectra are required. Based on such engineering problem, this study proposed a method to generate the seismic time history compatible with the multi-damping floor response spectra. In this method, the Hilbert-Huang Transform (HHT) is used to modify the frequency and amplitude information of seed motions, yielding the spectrum-compatible seismic time histories that completely satisfy the requirements of USNRC standards. Consequently, this study provides suitable seismic excitation for seismic safety evaluation of nuclear equipment and components.

Transient Structural Analysis of Fuel Assembly for Nuclear Powered Ships

Liang Shuangling, Wu Wanye

2019, 40(6): 89-94. doi: 10.13832/j.jnpe.2019.06.0089

Abstract(257)

Abstract:
In order to analyze the structural safety of fuel assembly for nuclear powered ships in moving state, taking the marine nuclear power platform as an example, hydrodynamics is adopted to conduct the frequency and time domain calculation for the platform to obtain the time serial curves of the center of the gravity in six degrees of freedom. Remote displacement method is used to transfer the ship motions to the reactor to realize the numerical simulation of fuel assembly moving with the ship, and transient structural mechanics is adopted to calculate the structural loads of fuel assembly finally. Results show that the structural loads of fuel assembly increase obviously in ship moving state compared with that in static state. Therefore, it is quite essential to take the ship random motion response into account during the structural safety analysis of fuel assembly.

Design Improvement of DBA Environmental Test Facility for Generation Ⅲ Nuclear Power Plants

Xing Limiao, Zhan Li, Xie Xu, Wu Xiaofei, Xu Shijie, Xu Changzhe, Ma Nan

2019, 40(6): 95-99. doi: 10.13832/j.jnpe.2019.06.0095

Abstract(710)

Abstract:
The environmental qualification test facility of Nuclear Power Institute of China, cannot meet the requirement of the environmental test of loss of coolant accident (LOCA) in the third generation NPPs, for the issues such as the inadequate capability of temperature rising and the uneven temperature distribution in the test chamber during the thermal shock. To solve these issues, the RELAP5 code was used for the numerical analysis for the design improvement of the test facility. Based on results of the numerical analysis, the facility was improved. The test results from the improved facility demonstrated that the capability of temperature rising in the test chamber during the thermal shock and the uniformity of the temperature distribution was largely improved. The new facility meets the requirements of design base accident (DBA) environmental test for the third generation NPPs.

Study on Heat Transfer Capability of Passive Residual Heat Removal System Based on Heat Pipe

Xian Lin, Zhou Ke, Li Feng, Yang Fan, Zhang Zhuohua, Zhang Dan, Wang Xiaoji

2019, 40(6): 100-104. doi: 10.13832/j.jnpe.2019.06.0100

Abstract(223)

Abstract:
Separated heat pipe has been applied in the industrial field, and is mostly applied in energy-saving field such as waste heat recovery boilers, because of its advantages such as fully separation of hot surface and cooling surface, completely isolation of heat source and cold source, and high heat exchange efficiency. Based on the working characteristics of the separated heat pipe, the passive residual heat removal system based on the heat pipe was developed. The theoretical design of the residual heat removal system is analyzed and demonstrated based on the system program. The parameters on the performance of the heat pipe is obtained including the different liquid charging rates, the elevation difference between heat source with cold source and the condensation section area ratio. When the liquid charging rates is larger than 0.7, the heat transfer power of the system decreases obviously. Even without the elevation difference between heat source with cold source, the system still can build the convection. The heat transfer power of the system increases proportionally with the increasing of the condensation section area ratio, when it ranges from1.0 to 2.5.

Prediction of Reactor Power under Accident Conditions of Nuclear Power Plant Using ν-Support Vector Machine

Jiang Botao, Huang Xinbo, Hines J.Wesley, Zhao Fuyu

2019, 40(6): 105-108. doi: 10.13832/j.jnpe.2019.06.0105

Abstract(127)

Abstract:
Aiming at the characteristics of core power change under accident conditions and the problems of artificial neural network (ANNs) such as easy to trap minimum and slow convergence speed, a prediction method of core power under accident conditions based on support vector regression (ν-SVR) was proposed. This method used a k-CV to optimize the parameters of ν-SVR, and then two different ν-SVR predictors were designed. These two predictors were applied to the prediction of core power of rod ejection accident (REA) and rod drop accident (RDA). The results have shown that this method has higher prediction accuracy and shorter response time than the ANNs.

Study on Power Control of Liquid Molten Salt Reactor Based on Core Linearization Model

Du Shangmian, Zeng Wenjie, Yu Tao, Chen Lezhi, Xie Jinsen

2019, 40(6): 109-113. doi: 10.13832/j.jnpe.2019.06.0109

Abstract(208)

Abstract:
The liquid molten salt reactor uses liquid molten salt as fuel, and the fuel flows in the primary loop driven by a main pump, resulting in reactivity loss in the process of flow. Considering the effect of fuel flow on core power control, a nonlinear model is established and linearized. Based on the core linearization model, the LQG/LTR technology is used to design the core power control system. Taking the molten salt experimental reactor as an example, the core reactivity disturbance control is studied. The results show that the LQG/LTR technology can be used to control the core power of liquid molten salt reactor.

PSA Analysis and Design Optimization for Reduction of Medium Safety Injection Pump Shut off Head of HPR1000

Yang Jian, Deng Chunrui, Ma Chao

2019, 40(6): 114-117. doi: 10.13832/j.jnpe.2019.06.0114

Abstract(618)

Abstract:
Through the probability safety analysis(PSA), it is found that in the design process of HPR1000, the reduction of the medium safety injection pump shut off head is beneficial to meet the steam generator tube rupture(SGTR) accident acceptance criteria, but it increases the core damage frequency (CDF) of loss of direct current accident. This paper puts forward the optimization scheme to add the rapid pressure relief valves for feed and bleed in the accident procedure. Applying this scheme to HPR1000, the PSA results show that the CDF of loss of direct current accident decreases from 2.4×10-8 (reactor?year)-1 to 2.2×10-9(reactor?year) -1. So the optimization scheme proposed in this paper effectively reduces the risk of HPR1000.

Application Analysis of SSG-30 of International Atomic Energy Agency(IAEA) Safety Classification Principle

Si Hengyuan

2019, 40(6): 118-123. doi: 10.13832/j.jnpe.2019.06.0118

Abstract(1663)

Abstract:
The purpose of the safety classification is to ensure that items are designed, manufactured, constructed, tested and operated with appropriate requirements to ensure that items have the appropriate quality under all expected operating conditions and to ensure the realization of safety functions. The Safety Classification of Structures, Systems and Components (SSC) in Nuclear Power Plants (SSG-30) was issued in 2014 by the International Atomic Energy Agency (IAEA). The principle of the safety classification covers five defense levels of nuclear power plants. It identifies the importance of safety-related items from two dimensions of design provisions and function categorized, considers the requirements of operating conditions and radioactive and operational limits of nuclear power plants, and determines the safety classification of items and related code requirements.

Online Identification of Regulator Model Based on Adaptive Forgetting Factor RLS Algorithm

Qian Hong, Jiang Cheng, Pan Yuekai, Shi Zhefeng

2019, 40(6): 124-129. doi: 10.13832/j.jnpe.2019.06.0124

Abstract(198)

Abstract:
In order to improve the accuracy of the time-varying system model identification of the voltage regulator and the rapidity and robustness of the on-line identification of the parameters, and to study the effect of forgetting factor on the performance of forgetting factor recursive least squares algorithm, an adaptive forgetting factor recursive least squares algorithm based on fuzzy algorithm is proposed in this paper. The average value of time series of the residual and its change rate between the dynamic characteristic value of the system and the identified model value are taken as the input of the fuzzy algorithm, and the correction of the forgetting factor is taken as the output to realize the adaptive adjustment of the forgetting factor. The simulation results of a regulator pressure reduction system of a nuclear power plant show that the algorithm can adjust the forgetting factor in real time, and effectively solve the time-varying problem of the parameters of the regulator model. Thus it can obtain a more accurate time-varying model, and can effectively solve the contradiction between the stability and the convergence speed of parameters identification results. Therefore, the algorithm is feasible and superior.

Assessment of Filtering and Pressure Drop Performances of IRWST Sump Strainers for Nuclear Power Plants

Xie Honghu, Li Shilei, Zhang Feng, Chen Chuyuan

2019, 40(6): 130-134. doi: 10.13832/j.jnpe.2019.06.0130

Abstract(206)

Abstract:
In this paper, filtering property and pressure drop performances of the in-containment refueling water storage tank(IRWST) sump strainers system were studied. The research work was performed with experimentation and numerical simulation methods, including experimental research on the downstream debris concentration and the pressure drop analysis of the strainers. The results show that the downstream debris concentration is 368 ppm under accidental condition, and the pressure drop of the RIS & HER sump strainers is 3.533 kPa and 3.631 kPa, respectively, which meet the functional requirements of the downstream debris concentration (480 ppm) and system pressure drop (5.6 kPa) of the IRWST sump strainers.

Study on Thermoelectric Power Measurement to Thermal Aging Effect of Precipitation-Hardened Martensitic Stainless Steel

Xue Fei, Shi Fangjie, Sun Qi, Chu Yingjie, Ti Wenxin, Huang Fei

2019, 40(6): 135-139. doi: 10.13832/j.jnpe.2019.06.0135

Abstract(219)

Abstract:
The change of mechanical properties and thermoelectric power with the extension of ageing time on the type 17-4PH precipitation-hardened martensitic stainless steel at 400℃ was studied. An empirical formula for the relationship between thermoelectric power and impact toughness was put forward and verified by the main steam isolation valve stem that has been in service for 13 years in a nuclear power plant. Results indicated that with the extension of aging time, the impact toughness of the material decreases, and the yield strength, tensile strength and hardness increase, and the shrinkage of the section and elongation of the crack decrease. The change of the thermoelectric power of the material was exponential correlation with the impact toughness, and the yield strength, tensile strength, hardness and thermoelectric power of the material show a good linear relationship. Good agreement was observed between impact toughness evaluated by thermoelectric power measurement and laboratory test value.

Design and Application of Ultrasound Stress Wave Signal Bandpass Filter

Jiang Zhaoxiang, He Pan, Zeng Jie, Wang Lei, Wang Guangjin, Liu Caixue

2019, 40(6): 140-143. doi: 10.13832/j.jnpe.2019.06.0140

Abstract(159)

Abstract:
According to the features of ultrasound stress wave signal, adopting a simple and accurate web filter design software, a eight-order Butterworth bandpass filter with the frequency range between 50 kHz and 200 kHz is designed. Using the circuit simulation software (Multisim), the filter circuit is tested and verified, by continuously changing the value of resistance and capacitance and optimizing the filter performance. The simulation results meet the design demands, which proves the reliability of the design. Using Labview to establish the automatic test platform, improves the test efficiency and saves the labor cost. By applying the filter to the nuclear pipeline evaluation system for HPR1000, the sealing performance of the main pipeline and the surge pipeline is monitored in real time.

Technology and Management of Reactor Coolant Pumps Assembling in Pressurized Water Reactors

Zhu Wei, Wang Feng, Wang Yongge, Wang Yong

2019, 40(6): 144-148. doi: 10.13832/j.jnpe.2019.06.0144

Abstract(195)

Abstract:
Reactor coolant pump (RCP) is the key equipment of nuclear power plants. This paper describes the structure, process and key technology for the assembling of the RCP. The main factors affecting the progress and quality of the RCP assembling were found out by summarizing and analyzing the assembling problems. In addition, measures, such as strengthening the management of assembling plan, reasonably arranging the provision of labor, making good technical preparation before assembling and establishing emergency treatment channels, are taken to effectively improve the assembling quality and efficiency, which can provide experience reference for the RCP assembling in other nuclear power projects.

Structural Design and Finite Element Analysis of a Normally-Closed Finned Freeze Valve

Jiang Xinyue, Wang Naxiu, Su Bo, Chen Yushuang, Kong Xiangbo, Lu Huiju, Fu Yuan

2019, 40(6): 149-154. doi: 10.13832/j.jnpe.2019.06.0149

Abstract(172)

Abstract:
The freeze valve TMSR-FV1 functions as a normally closed shut-off valve in the secondary loop of the Thorium-based Molten Salt Reactor (TMSR). TMSR-FV2 is an improved version of TMSR-FV1, which heat transfer is enhanced by adding fins to the outer surface of freeze valve while keeping its size intact. In this paper, the Finite Element Method was used to make structural design and thermal analysis of TMSR-FV2. The effects of fin height, fin thickness and fin pitch on the cooling of molten salt in TMSR-FV2 under natural cooling conditions were studied. A finned freeze valve structure was proposed and compared with the experimental salt plug effect of TMSR-FV1 under natural cooling and different cooling flow conditions. The calculation results show that the proposed TMSR-FV2 effectively reduces the central molten salt temperature of the flat section to enhanc the salt plug of the freeze valve and realizes the passive closing function of the freeze valve within the scope of the research.

Experimental Study on Design and Layout of Recirculation Valve for Main Feedwater System of Nuclear Power Plants

Mu Guanyu

2019, 40(6): 155-158. doi: 10.13832/j.jnpe.2019.06.0155

Abstract(95)

Abstract:
The design and arrangement of the recirculation valve in the main water supply system of a nuclear power plant is studied, to explore the causes for the sudden jump of the recirculation pipeline at the moment of starting the pump and accompanied by the blasting sound. The results show that the multi-stage cage regulating valve cannot be arranged in the high-pressure water supply pipeline with residual air, otherwise the destructive water hammer will be induced during the start-up stage. By optimizing the design layout of the recirculation valve, the problem of abnormal startup of the main water supply system was finally solved.

Analysis and Solution for Effect of IRC Premature Saturation on ATWT Function

Zheng Junwei, Liu Yang, Zhang Juanhua, Liu Zhaopeng, Niu Maolong

2019, 40(6): 159-162. doi: 10.13832/j.jnpe.2019.06.0159

Abstract(198)

Abstract:
The output current of intermediate range channel (IRC) of nuclear instrumentation system (RPN) reached the saturation value prematurely when the nuclear power (Pn) was less than 30%FP during the commissioning startup of a pressurized water reactor nuclear power plant. The main influence factor of IRC premature saturation was its neutron detector manufactured sensitivity (SF) , after the first cycle refuel design scheme (FCRS) determined. It can be concluded that IRC premature saturation is the reason that the actual threshold of ATWT permit signal is lower than its design expected threshold, based on the ATWT design function and nuclear instrumentation system (RPN) safety criteria. In order to ensure the correct trigger of ATWT permit signal, the threshold Pnof ATWT permit signal has to be selected as Pn under the saturation power when IRC premature saturation occurred. In order to decrease the probability of the IRC premature saturation, the IRC detector with smaller SF is sugguested for the nuclear power plants with FCRS of 18 months period (FCRS18) .

Analysis and Optimization of Frequent Isolation of Steam Generator Blowdown System in CPR1000 Nuclear Power Plant

Xu Ying, Xu Jinquan, Yang Zongwei, Yu Hang

2019, 40(6): 163-167. doi: 10.13832/j.jnpe.2019.06.0163

Abstract(347)

Abstract:
The steam generator blowdown system (APG) of CPR1000 nuclear power plant is frequently automatically isolated during normal operation. In combination with the operating background of the isolation events, the causes of these events are studied, and APG is optimized. The D link is added to the cooling water temperature controller and the control parameters are optimized; the feedforward link is added to the sewage flow control loop; the corresponding delay links are added to the downstream of sewage flowmeter and pressure switch signal; the adaptive modification is made to the operation program of the start-up stage. The actual operation of the nuclear power plant has proved that the improved system is in better operation condition, the automatic isolation events are greatly reduced, and the operation and maintenance costs are effectively lowered.

Design of a Ventilation Flow Orifice Plate Aperture Regulator

Wang Xiaoming, Yang Hongwei

2019, 40(6): 168-172. doi: 10.13832/j.jnpe.2019.06.0168

Abstract(151)

Abstract:
For the commissioning of the WWER nuclear island ventilation system, the fixed flow orifice plate is used for the air volume adjustment, and the on-site operation is heavy and time-consuming. By analyzing the principle of the aperture mechanism, a ventilation flow orifice plate aperture adjuster is developed. The test results show that the regulator realizes the function of changing the diameter of the line to adjust the air volume, and has the characteristics of easy installation and high precision. The device is used for the system commissioning, avoiding the repeated changes of the fixed-diameter orifice plates, which is beneficial to improve the commissioning efficiency, and also provides reference for the design of other stack-type ventilation systems.

Calculation and Analysis of Shielding Effectiveness and Radio-Toxicity of Depleted Uranium Applied to Nuclear Power Reactor Protection

Song Yingming, Wang Yan, Xiao Feng, Lyu Huanwen, Fu Mengting, Shen Geyu

2019, 40(6): 173-177. doi: 10.13832/j.jnpe.2019.06.0173

Abstract(217)

Abstract:
In this paper, the gamma ray shielding of depleted uranium is verified by experiment and simulation. A nuclear power pressurized water reactor shielding model is constructed, and the simulation distribution of neutron energy spectrum in the shielding layer was consistent with that in practice. The comprehensive shielding performance of depleted uranium in different positions of neutron and gamma mixed radiation field were simulated by the Monte Carlo simulation method and the fuel consumption calculation program, and compared with that of the lead shielding material. The activation and fission products of the depleted uranium irradiated by the shielding neutrons for 40a was simulated and calculated, and the radio-toxicity defined by Annual Limit of Intake (ALI) before and after irradiation of the material was analyzed. The results show that the addition of secondary products has little effect on the radio-toxicity.

Design and Research of a Novel RVI with Separated Barrel Hang in Middle of RPV for ACP100

Zhang Hongliang, Fan Heng, Xu Bin, Wang Liubing, Chen Xungang, Xu Haibo, Liu Xiao, Li Hao

2019, 40(6): 178-182. doi: 10.13832/j.jnpe.2019.06.0178

Abstract(154)

Abstract:
A novel reactor vessel internals (RVI), which a barrel assembly is fixed by the compression barrel assembly, and hang in the middle of reactor pressure vessel (RPV), was proposed by analyzing the structure characteristics of China small modular reactor (ACP100). This structure has many advantages, such as easy for manufacture and installation, good sealing performance, and well flow distribution capability. The results of mechanical analysis and control rods drive line comprehensive experiment and flow-induced vibration test show that it is with high security and reliability, and fully meets the functional requirements of ACP100.

Study on Mitigation Strategies for ACP600’s Main Steam Line Break Accident

Zhang Shu, Qiu Zhifang, Zhang Xiaohua, Chen Hongxia, Fang Hongyu

2019, 40(6): 183-188. doi: 10.13832/j.jnpe.2019.06.0183

Abstract(156)

Abstract:
The main steam line break (MSLB) accident of ACP600 advanced third generation nuclear power plant is taken as the research object. The MSLB accident of ACP600 is analyzed and studied aiming at the disadvantageous situations that ACP600 cancels high pressure safety injection system and boron injection tank, using integrated gadolinium as burnable poison, adopting mode-Coperation and control mode and so on. For MSLB accident, because of the high peak power of the reactor core after re-reaching the critical state, in order to avoid the minimum DNBR(Departure from Nucleate Boiling Ratio) below the limiting value, the effects of different injection system configurations and tripping the fault loop’s reactor coolant pump on mitigating MSLB accident are evaluated from the viewpoints of rapid injection of boron solution and reduction of core cooling rate, and the best mitigation scheme is studied, the design optimization advice about adding "steam line pressure low-3 " signal for shutting down the fault loop is put forward.

Study on Applicability of RELAP Code in PRS Analysis

Zhang Xiaohua, Li Feng, Zhang Yu, Wu Qing, Yu Na, Zhang Shu, Xian Lin

2019, 40(6): 189-193. doi: 10.13832/j.jnpe.2019.06.0189

Abstract(234)

Abstract:
In this paper, the passive residual heat removal system (PRS) on secondary side of HPR1000 nuclear power plant is taken as the research object, and the suitability of applying RELAP code to PRS analysis is evaluated. Firstly, by identifying the key phenomena and sorting out program functions, it shows that RELAP code related mathematical model can meet the needs of PRS analysis. At the same time, the applicability of the model was quantitatively evaluated by modifying the heat transfer model of steam condensation in horizontal/vertical tubes in RELAP program and comparing with the results of PRS test bench. On the basis of the above work, the transient process of PRS operation under SBO condition is analyzed by using RELAP. The results show that PRS can establish natural circulation and effectively discharge heat. Finally, the transient condition of PRS misoperation is analyzed. The results show that the reactor is safe under the misoperation condition of PRS, and the design of PRS system meets the requirements.

Development of Point-Depletion Code Based on Chebyshev Rational Approximation Method

Zhang Yunfei, Zhang Qian, Zhao Qiang, Zhang Zhijian

2019, 40(6): 194-197. doi: 10.13832/j.jnpe.2019.06.0194

Abstract(224)

Abstract:
A point-depletion code is developed based on the Chebyshev Rational Approximation Method (CRAM). The code has two burnup libraries, a detailed burnup library and a simplified burnup library. The code is coupled with the transport system. The fixed flux irradiation and decay problem, as well as the JAEA benchmark for light water reactors, were calculated and compared with international well-known codes. The results show that the accuracy of the atom number density is equivalent to that of ORIGEN2 in the calculation of fixed flux irradiation and decay problem. The results of pin cell and assembly cases agree well with HELIOS1.11 and the reference solutions.

Research on Multi-Objective Optimization Design Method for Nuclear Power Systems

Zhu Li, Peng Shinian, Yang Yunjia, Li Feng, Xian Lin, Zhang Dan, Qiu Zhifang, Yuan Hongsheng

2019, 40(6): 198-202. doi: 10.13832/j.jnpe.2019.06.0198

Abstract(208)

Abstract:
The modified multi-objective self-adaptive differential evolution algorithm is proposed to improve the precision of the multi-objective optimal design results. MMOSADE is confirmed by the ZDT test function set. A multi-objective optimization design method for nuclear power systems is developed. The method is applied to the PRHR-HX of AP1000. The optimal results indicated that this optimization design method is effective and feasible.

Measurement of Rectangular Channel Bubble Fraction Based on Wire-Mesh Technology

Yang Yiang, Xiong Jinbiao, Zhang Tengfei, Chai Xiang, Liu Xiaojing

2019, 40(6): 203-206. doi: 10.13832/j.jnpe.2019.06.0203

Abstract(413)

Abstract:
Based on the wire-mesh sensor technology, a novel non-invasive sensor for narrow channel void fraction measurement has been developed, by changing the structure of the wire-mesh sensor (WMS). Through two-phase experiment, the feasibility of this sensor has been verified by compared with high speed camera. In addition, a corresponding post-processor has been developed. To improve the accuracy of the post-processor, based on high speed camera, we evaluate the algorithms used by post-processor. As a result, lanczos interpolation algorithm and Canny edge extraction algorithm were finally selected.

2019 Vol. 40, No. 6