核动力工程

Thoughts on Decommissioning of Nuclear Power Plants in China

Wang Xin, Wei Shuhong, Dai Bo

2019, 40(3): 1-5. doi: 10.13832/j.jnpe.2019.03.0001

Abstract(256)

Abstract:
With the increasing of the numbers and operation years of the nuclear power plants in China, the decommissioning of nuclear power plants is becoming more and more important. The safe, economical and environment-friendly decommissioning is related to the sustainable development of the nuclear industry, as well as the safety of the ecological environment and the health of future generations. From the perspective of international nuclear decommissioning experience feedback and decommissioning market development and based on decommissioning related national planning in China, key tasks in the future decommissioning of nuclear power plants in China are proposed, including the completion of the decommissioning related codes and standard; Self-reliant, enterprization, marketization, and specialization road, cooperation and win-win strategy; the decommissioning related considerations of the whole life cycle of nuclear power plants, the acceleration of the R&D of decommissioning related key technology and equipment, and the provision of decommissioning funds.

Prediction of Critical Heat Flux in Non-Uniformly Heated Tubes Based on Two-Phase CFD

Li Quan, Chen Ping, Huang Yongzhong, Chen Jie, Jiao Yongjun, Yu Junchong, Maria Avramova

2019, 40(3): 6-11. doi: 10.13832/j.jnpe.2019.03.0006

Abstract(979)

Abstract:
In order to establish the critical heat flux (CHF) prediction method with non-uniform heat flux, as an additional approach for the safety analysis in the heat transfer systems, Eulerian two-fluid model coupled with wall boiling model are used to predict the critical heat flux in non-uniformly heated tubes. The calculated wall temperature and near-wall void fraction distributions under different heat fluxes are obtained and compared. The second peak of wall temperature and near wall void fraction are used as the criteria for CHF and the location of the highest near wall void fraction is regarded as the location of CHF. Two different power distributions are researched. The prediction has good agreement with the experiment, including both the CHF and their locations. Thus, the prediction method used in this paper can be used in the CHF prediction in non-uniformly heated tubes.

Numerical Investigation on Flow and Heat Transfer Characteristics of S-CO2 in Narrow Space Channel of PCHE Based on CFD Simulation

Liu Guijun, Chen Deqi, Hu Lian, Wang Junfeng

2019, 40(3): 12-16. doi: 10.13832/j.jnpe.2019.03.0012

Abstract(1077)

Abstract:
CFD method was used to investigate the thermal hydraulic characteristics of the printed circuit heat exchangers (PCHE) with zigzag channels, and the physical parameters of supercritical CO2 was calculated by combining the physical property fitting formulas and the user defined function (UDF ). In addition, the turbulent flow model was based on the SST k-ω model, and the CFD method was validated by the experimental data of friction coefficient and heat transfer coefficient. Simulation results show that in the zigzag channels of PCHE, the core accelerating area moves toward the outside wall at the bending point, and the separated flow area becomes narrow as the pressure drop decreases at the downstream part of the bending point.

Test Verification of Wave-Plate Separation Element for Moisture Separator Reheater

Zhang Qian, Luo Huan, Feng Jing, Wang Yong, Wang Wei, Wei Li

2019, 40(3): 17-20. doi: 10.13832/j.jnpe.2019.03.0017

Abstract(646)

Abstract:
The separation characteristics and pressure drop characteristics of the saturated wet steam through the corrugated plate separation element are verified under the actual operating conditions of the moisture separator reheater (MSR) for PWR nuclear power units. The saturated wet steam which meets the requirements are provided by using a large thermal hydraulic test bench, and the pressure drop and residual humidity are obtained when the wet steam flows through the test piece. The results show that the critical flow rate of re-entrainment is high, and when the inlet steam dryness is not greater than 86% and the inlet flowrate is less than the critical flowrate, the residual humidity is not greater than 0.1%, the separation efficiency is of 99% or more, and the pressure drop is not greater than 7 kPa. The design requirements of more than 99% steam dryness and less than 14kPa pressure drop are satisfied. It is with good separation and pressure drop characteristics.

Numerical Simulation on Flow and Heat Transfer Characteristics of Sodium in an Annuli

2019, 40(3): 21-25. doi: 10.13832/j.jnpe.2019.03.0021

Abstract(803)

Abstract:
Numerical simulation on the flow and heat transfer characteristics of sodium flowing in an annuli was conducted using ANSYS CFX. The inside diameter of the annuli was from 4mm to 10 mm while the outside diameter from 7mm to 20 mm. The influence of the velocity, inlet temperature, heat flux from the inner surface and outside surface were analyzed. Calculation results showed that the gap of the annuli was with great effects on the flow and heat transfer characteristics of sodium significantly, but the heat flux with little effect. Results in this work were compared with the theoretical and experimental results from literatures. A correlation of the heater transfer coefficient for liquid sodium flowing in an annuli was proposed based on the analysis. The calculation results in this work show good consistency with other theoretical and experimental results.

Parallel Technology of Fluid-Solid Heat Transfer in Full Scope Simulator of Nuclear Power Plants

Chen Junjie, Zhao Xiumei, Lin Meng, Wang Xu, Li Yankai, Chen Gang

2019, 40(3): 26-31. doi: 10.13832/j.jnpe.2019.03.0026

Abstract(724)

Abstract:
Based on the theory of thermal hydraulic calculation of RELAP5, the application of the coupled parallel technology by heat flux and wall temperature in the full scope simulator of nuclear power plants is studied. The model splits up into two coupling parallel computing models by wall temperature or heat flux of the heat structure. Comparing the results with the overall calculation results, it shows that: the results of coupling calculation by heat flux of recuperative heat exchanger are inaccuracy; and the results of coupling calculation by wall temperature are consistent with the actual value. The simulation results of the steam supply system of the typical nuclear power plant by using coupling parallel computing model by wall temperature shows that: the coupling parallel computing by wall temperature of recuperative heat exchanger effectively improves the CPU utilization and the computation speed, which provides more guarantee for the real-time calculation of the simulation.

Iterative Method of Flow Computation in Guide Thimble during Rod Dropping

Liu Sheng, Yang Yiren, Li Peng

2019, 40(3): 32-37. doi: 10.13832/j.jnpe.2019.03.0032

Abstract(732)

Abstract:
In the previous rod dropping computation, it was assumed that the flow in the discharge orifices of the guide thimble are all of outflow throughout the rod dropping course, while, in fact, the inflow occurs. To handle this problem, it is assumed that inflow and outflow exist in each discharge orifice. The energy dissipation is analyzed on every flow section, having changes in shape or area. By introducing the flow indexes to represent the flow status of each orifice, a set of equations of the Bernoulli and flow conservation is subsequently developed by accumulating all flow status with these indexes. The flow status, pressure and velocity in the guide thimble are iteratively solved and used to compute the drag force on the control rod assembly. The dropping time is finally solved by substituting the drag force into the equation of dropping motion. The dropping time is obtained as 2.133 s while it is 2.036 s if the flow reversed is neglected and the outflow is assumed. It is indicated that the flow reverse has a significant effect on the dropping course and dropping time. Hence, this iterative method is a preferable choice of the flow computation for the guide thimble.

Experimental Research on Water Level Measurement Characteristics of Pressurizer in Transient Condition

Liu Yan, Wang Yanzhi, Liang Linke, Wang Xianyuan, Tian Ruifeng

2019, 40(3): 38-42. doi: 10.13832/j.jnpe.2019.03.0038

Abstract(762)

Abstract:
Pressurizer is the main equipment of the pressure safety system of PWR nuclear power plants. Its water level reflects the water volume of the primary loop, which is one of the key parameters for the operation and control of the pressurizer. Based on the two-zone unbalanced model, this paper simulates the water level change under the condition of steam leakage and tests the characteristics of water level measurement for the steam leakage condition of the pressurizer, and the pressure drop of the water level under the pressure drop rate of 2.6~7.8 kPa/s in the pressurizer was investigated. It is found that the water level calculated by the pressure difference correction liquid zone density has a good follow-ability in the case of pressure transient, which can better reflect the water level characteristics. Pressure drop characterizing the change of the density in the liquid region of the pressurizer during the pressure reduction, the transition time is less than 40s and there is no strong correlation between the transition time and the single factor of the pressure drop rate. This provides basic research data for the safe operation of the pressurizer.

Research and Improvement of Diagnosis Method of Fuel Failures for Pressurized Water Reactor Nuclear Power Plant

Lyu Weifeng, Xiong Jun, Chen Xiaoqiang, Tang Shaohua, Liu Jie

2019, 40(3): 43-47. doi: 10.13832/j.jnpe.2019.03.0043

Abstract(998)

Abstract:
The diagnosis methods in the world for the fuel failures in PWR and the existing problems of diagnosis methods are analyzed and improved from three aspects, which are the number of failed fuel rods, fuel failure size and burnup of failed fuel. The factors that may affect the diagnosis result are also identified and discussed. The operating data of Chinese in-service nuclear power plants are used to verify the improved diagnosis method, and the results show that, the improved diagnosis method for fuel failures can accurately diagnose the fuel failure and has a wider applicability.

Calculation Study of Prompt Dose from Uranium Processing and Fuel Fabrication Facility in Criticality Accident

Shen Haibo, Liu Aihua, Hu Wei, Xiao Hongwen, Huang Dan

2019, 40(3): 48-51. doi: 10.13832/j.jnpe.2019.03.0048

Abstract(696)

Abstract:
A method for calculating the prompt dose of Uranium Processing and Fuel Fabrication Facilities under nuclear criticality accidents based on MCNP code modeling was established, and the analysis and comparison was done with the calculation method stipulated in criteria EJ/T988-96. Then took a fuel research and development site at home as an example, the prompt dose of public on the edge of site was calculated. The results showed that the prompt dose in a nuclear criticality accident was overvalued using the calculation method stipulated in criteria; The modeling method based on MCNP made the calculation result more accurate because of its scientific algorithm, accurate description of shielding medium and controllability of the result error. Therefore, it is suggested that the MCNP-based code modeling method be used to assess the prompt dose under the nuclear criticality accidents of Uranium Processing and Fuel Fabrication Facilities.

Research on Shielding Performance of Dry Storage Cask for Spent Fuels

Huang Liming, Tang Shaohua, Liu Jie, Yang Shouhai

2019, 40(3): 52-56. doi: 10.13832/j.jnpe.2019.03.0052

Abstract(726)

Abstract:
The radiation shielding design of spent fuel storage containers is one of the important factors affecting the radiation safety of spent fuel dry storage facilities. Taking the dry storage container independently researched and developed by China as an example, the paper focuses on the reasonable determination of the shielding performance target, the shielding calculation method, the selection of the calculation hypothesis and the analysis of the calculation results. The calculation analysis shows that the designed spent fuel dry storage container is with good shielding performance and meets the radiation safety requirements.

Determination of Trace Uranium in Nuclear Grade Sponge Zirconium by Inductively Coupled Plasma Mass Spectrometry

Liao Zhihai, Long Shaojun, Wang Peng, An Shenping, Luo Fengyan, Qiao Hongbo

2019, 40(3): 57-60. doi: 10.13832/j.jnpe.2019.03.0057

Abstract(673)

Abstract:
In order to ensure the quantify control of the nuclear grade zirconium sponge, the inductively coupled plasma mass spectrometry(ICP-MS) was used to determine the trace uranium in the nuclear grade sponge zirconium. The sample was dissolved quickly by HNO3-HF solution. The analysis of the choices of isotope mass number and internal standard isotope, and the influence of matrix showed that the effective correction of matrix effect and the interference caused by the instrument drift can be accomplished, while 232Th was used as an internal standard. When the uranium standard stock solution with concentrations of 0.20 μg/g, 1.00 μg/g and 1.50 μg/g were respectively added to the samples, the recovery rates were between 94% and 99%, with corresponding maximum Relative Standard Deviation(RSD) of 5.1%, relative standard uncertainty of 5.9%. The detection limit was 0.001 μg/g. This method was employed to determine the actual sample, and the results meet the requirement of the uranium content in the nuclear grade sponge zirconium.

Parameters Measurement of Nuclear Graphite Based on Digital Image Correlation

Yi Yanan, Zhang Xiaojuan, Ma Shaopeng, Zhu Haibin, Sun Libin, Shi Li, Jiang Han

2019, 40(3): 61-65. doi: 10.13832/j.jnpe.2019.03.0061

Abstract(792)

Abstract:
This paper develops a method for measuring the mechanical parameters of nuclear graphite based on digital image correlation (DIC) method, and gives detailed data processing method and measurement procedure. The experimental results show that the Young's modulus and Poisson's ratio extracted from DIC data are in good agreement with those from the strain-gauge-based method, which verifies the feasibility and accuracy of the DIC method. In addition, the complete stress-strain relationship was obtained from DIC method, confirming the superiority of DIC method in measuring the mechanical parameters of materials with large deformation. Furthermore, the DIC method is easy to operate and efficient in identify the loading misalignment, making it convenient in large number of tests and reliable for the data acquisition for brittle materials testing.

Sensitivity Analysis of Pressure Response in Containment Cooled by Natural Circulation

Sun Yanyu, Zheng Yuntao, Ma Xiuge, Yang Changjiang

2019, 40(3): 66-69. doi: 10.13832/j.jnpe.2019.03.0066

Abstract(733)

Abstract:
Passive containment air cooling system(PAS) is a very important system of modularized small nuclear power plants, which is the key part in the design of the containment. This system can ensure that the pressure is below the design limit and guarantee the integrity of the containment. This paper analyzes the sensitivity of 13 parameters that may influence the containment pressure response, using the Latin hypercube sampling(LHS). The results showed that the initial temperature in the containment and the temperature of the environment are most sensitive to the pressure response. In this paper, the sensitivity analysis of the containment pressure is carried out by using a general best estimate software and a statistical analysis method in the first time. This result can support the design of the containment, the safety analysis and the safety review.

Kinetic Simulation of Lifting Mechanism of Personnel Airlock in Nuclear Power Plants

Xie Honghu, Ma Wenqin, Zhang Feng, Yang Jinchun

2019, 40(3): 70-74. doi: 10.13832/j.jnpe.2019.03.0070

Abstract(661)

Abstract:
In order to ensure the rationality of the structural design for personnel airlock lifting mechanism in nuclear power plants, the dynamic characteristics of the sealing door during the opening and closing process are analyzed. This paper adopts Automatic Dynamic Analysis of Mechanical Systems(ADAMS) software to analyze the airlock dynamics and kinematic performance, including the kinematic simulation of the driving box and the dynamic simulation of the transmission screw. The analysis indicates that the initial installation position of the driving box in the lifting mechanism is the key factor leading to the insufficiency of lifting height. By resetting the initial installation position of the driving box and revolving the velocity and runtime of the screw, the lifting height can meet the design requirement (60mm). Also, the dynamic simulation results of the screw show that the radial force imposed on the screw can be neglected, which would not cause the rupture of the screw.

Sensitivity Analysis of Reactor System Dynamic Responses

Huang Qian, Xiong Furui, Lan Bin, Wang Bihao, Song Haiyang, Huang Xuan, Liu Shuai

2019, 40(3): 75-80. doi: 10.13832/j.jnpe.2019.03.0075

Abstract(683)

Abstract:
The analysis of the sensitivity of the input parameters to the dynamic responses of the reactor system is an important prerequisite for optimal design. In this paper, the contact stiffness and clearance of the key parts of the reactor system are taken as input variables, and the Sobol’s method is used to analyze the sensitivity of the key input parameters to the dynamic response of the system under the seismic condition. The global sensitivity coefficient and the order of importance of the input parameters are obtained. The results were also tested using the K-S measure sensitivity analysis method. The analysis shows that the seismic response of the fuel assembly is sensitive to the change of the contact stiffness at two locations, and the tangential load extreme value is most sensitive to the change of the contact stiffness at the location. The relevant methods and analysis processes can be extended to the reactor coolant system and other main equipment to provide the method for quantitative analysis and data support for the selection of optimized design parameters.

Design Technology for Highly Reliable Main Controller of Security Level in Nuclear Power Plants

Ma Quan, Luo Qi, Song Xiaoming, Liu Yanyang

2019, 40(3): 81-86. doi: 10.13832/j.jnpe.2019.03.0081

Abstract(754)

Abstract:
Considering the characteristics of the main controller of the digital control system (DCS) platform of security level and its function and application, based on the specifications in relevant laws, regulations and standards, studies had been conducted on the diagnosis, redundancy, communication and embedded design of the highly reliable main controller. Key technologies such as software development were studied and applied in the design of the main control module of the China National Nuclear Corporation's DCS platform of security level, the Nuclear Advanced Safety Platform of I&C (NASPIC). At the same time, the HPR1000 simulation platform was built and its functions and performance were tested in the shutdown function test, engineering safety facility actuation function test and periodic tests. These tests and the qualification test conducted by the National Nuclear Safety Administration showed that the diagnostic coverage rate reached 98%, exceeding the standard requirements; the measured communication error rate was less than 10-11, meeting or even exceeding the security level DCS product specifications of other manufacturers; the hot standby redundancy architecture and embedded software meet the requirements of Class 1E equipment, ensuring the high reliability of the main controller.

Predictive Control of Pressure non-Stabilization System of Pressurizers in Nuclear Power Plants

Qian Hong, Yuan Yuan

2019, 40(3): 87-92. doi: 10.13832/j.jnpe.2019.03.0087

Abstract(651)

Abstract:
Considering the features of the pressurizer pressure, such as large inertia, multiple disturbances, complex nonlinearity and difficulty in obtaining accurate mathematical models, and in view of the dynamic characteristics of unstable open loop of the pressurizer pressure, a predictive control system is devised in this paper. Firstly the dynamic matrix predictive controller is designed with the model self-stabilization of the non-stabilization system, and in order to solve the engineering implementation of this control method, the control signal output is obtained. Finally the control system simulation platform is built on MATLAB/Simulink. Simulation and disturbance test show a superior control performance and strong anti-disturbance ability. It provides a solution for the application of dynamic matrix control in the non-stabilization system.

nvestigation of Single-Step Reaction Mechanism for Hydrogen Combustion in Severe Accidents of Nuclear Power Plants

Zhu Yonghui, Wang Ying, Li Yong, Tang Yueming, Zheng Hua

2019, 40(3): 93-96. doi: 10.13832/j.jnpe.2019.03.0093

Abstract(745)

Abstract:
Changing the pre-exponential factor and the activation energy coefficient, the single-step reaction mechanism for hydrogen combustion was established, and the calculation and analysis of hydrogen combustion have been conducted by this mechanism. The results from the calculation and the experiment were compared, and the calculations of hydrogen combustion were processed in different hydrogen concentration by the single-step reaction mechanism. The results showed that the calculations and experimental flame propagation were conformable, and the modified single-step reaction mechanism for hydrogen combustion can be used to calculate and analyze the hydrogen combustion in nuclear power plants.

Numerical Analysis of ACME Station Blackout Integral Effect Test with Relap

Liu Yusheng, Xu Chao, Hu Jian, Zhuang Shaoxin, Fang Fangfang

2019, 40(3): 98-102. doi: 10.13832/j.jnpe.2019.03.0097

Abstract(304)

Abstract:
To study the performance characteristics of AP type passive nuclear power plant under typical station blackout (SBO) condition, numerical analyses are investigated based on the SBO integral tests conducted on the Advanced Core-cooling Mechanism Experiment (ACME) facility. The main parameters, which are obtained by establishing model with Relap5 code, are compared with that of SBO integral tests. The results show that the Relap5 model represents the major accident progress of ACME SBO tests well. The accidents chronology and key thermal hydraulic phenomena of post-calculations are consistent with that of tests. For natural circulation process between core and passive residual heat removal heat exchanger (PRHR HX), the calculated circulation rate is greater than that of tests, which results in a faster natural circulation process. The comparison analysis also shows that the results predicted by Relap5 are conservative for transient pressure of reactor coolant system (RCS) and maximum water level in pressurizer. There is a margin between numerical values and experimental values.

Reliability of Digital Pressure Control Device of Nuclear Pressurizer Based on Dynamic Fault Tree

Qian Hong, Gu Yaqi, Liu Xinjie

2019, 40(3): 103-108. doi: 10.13832/j.jnpe.2019.03.0103

Abstract(1082)

Abstract:
The digital pressure control device for the pressurizers in nuclear power plants, which is equipped with the input module of selecting median from four inputs, was used as the research object, and a fault tree including the static part and the dynamic part of the device is established. Next, the Method of Markov was used to analyze the dynamic part, and then the overall fault tree was analyzed according to the overall logic. Finally the reliability of the input module and its effect on the entire device were evaluated from the value of reliability and importance. The evaluation concluded that the input module is relatively highly optimized in terms of reliability.

Settlement Calculation and Prediction for Nuclear Projects Located on Soft Rock

2019, 40(3): 109-114. doi: 10.13832/j.jnpe.2019.03.0109

Abstract(587)

Abstract:
In order to discuss the deformation properties of the soft rock, based on the specific nuclear project located on the soft rock, a three dimension finite element model of Nuclear Island foundation was generated and the rheological constructive model which is suit for the soft rock was developed in this paper. This constructive model was embedded in software FLAC3D. The comprehensive assessment model of the attribute recognition based on the entropy weight of sensitivity was adopted in the sensitivity analysis for the geotechnical parameters of soft rock. And the inversion calculation applying the uniform design –advanced genetic analysis-neural network method was also carried out. Based on the monitoring data, the new geotechnical parameters used in elastic-plastic model and viscoelastic-plastic model were obtained through inversion calculation. Then the settlement analysis was carried out using the numerical method based on the optimized parameters. The settlement prediction also can be produced. The method used in this paper can be adopted in the settlement calculation of soft rock in the similar projects.

Study on Startup Control Characteristics for CSR1000

Yuan Yuan, Wang Li, Luo Hanyu, Shan Jianqiang, Zhang Xiaoying, Wang Dongqing

2019, 40(3): 115-120. doi: 10.13832/j.jnpe.2019.03.0115

Abstract(701)

Abstract:
The analysis of the startup system and startup characteristics are one important part of SCWRs design. In order to realize the analysis of the whole startup system, this paper proposes the new wall heat transfer model, based on the supercritical transient analysis code named SCTRAN. Then, the paper presents a control strategy for the startup process, including the control of coolant flow rate, core inlet temperature, system pressure, core thermal power, and steam drum water level. Different control schemes are set up according to different control objectives of the startup phases. Based on the CSR1000, an analytical model including the circulation loop and the once-through direct cycle loop is established, and four startup processes with control systems are put forward. The calculation results show that the thermal parameters of the circulation loop and the once-through direct cycle loop meet the expectation, and the maximum cladding surface temperature does not exceed the limit temperature 650℃. The feasibility of the startup scheme and the security of the startup process are verified.

Investigation on Vertical Temperature Profile Generated by Fire on Horizontal Multi-Layer Cable Tray in Closed Compartment of Nuclear Power Plant

Wang Yuhong, Zhu He, Huang Xianjia

2019, 40(3): 121-124. doi: 10.13832/j.jnpe.2019.03.0121

Abstract(742)

Abstract:
Focusing on the fire risk of the horizontal multi-layer cable tray in a nuclear power plant, the characteristics of the smoke layer temperature profile generated by the fire on the multi-layer cable tray in a closed compartment and the prediction model for the smoke layer temperature were investigated. Fire experiments of a three-layer cable tray with two typical cable arrangements were conducted in a closed compartment. The experimental results showed that the stratification of vertical smoke layer temperature was evident. Based on the vertical temperature profile in the compartment center, the thermal environment resulted from the fire can be classified as top ceiling jet layer, middle smoke layer, and lower cool air layer. The unsteady-state temperature prediction model in a closed compartment was used to predict the smoke layer temperature resulted from the fire on the horizontal multi-layer cable tray. The comparison of the predicted results with the experimental temperature for the smoke layer showed that this model can accurately predict the maximum temperature of the smoke layer, with relative error less than 1%. the prediction model underestimated the temperature generated by the decay stage of the cable tray fire, and thus the global error is within 16.3% and 27.8%.

Method of Developing Target Power Spectral Density Compatible with Design Response Spectra

Sun Yugang, Chu Meng, Huang Xiaolin, Ding Zhenkun

2019, 40(3): 125-129. doi: 10.13832/j.jnpe.2019.03.0125

Abstract(1269)

Abstract:
U.S. NRC SRP 3.7.1 requires that the time history for the SCCs seismic design of nuclear power plants should envelop both the design response spectra and the target power spectral density (PSD) function compatible with the design response spectra. According to SRP 3.7.1(2014), the method of developing the target PSD compatible with the design response spectra and the corresponding computer program was presented in this paper. The results of the computer program were verified by some examples. The results show that the calculated target PSD compatible with RG1.60, CEUS and WUS rock sites response spectra were consistent with the corresponding results provided in SRP 3.7.1, and the spectrum from the artificial time history which developed by the target PSD and the superposition method of trigonometric series adequately matched the design response spectra. The generation of target PSD in this paper can provide technical basis for the PSD check on the time history for the SSCs seismic design of nuclear power plants, and when the target PSD were developed using the method described in this paper, the PSD check should be set at 70 percent of the target PSD.

Reliability Growth Test and Improvement of Electromagnetic Movable Coil Control Rod Drive Line

Wu Xiaofei, Yang Xiaochen, Nie Changhua, Du Hua, Yan Xiao, Peng Hang, Yang Zumao, Tang Hui, Xiao Linhai, Ma Nan, Xing Limiao

2019, 40(3): 130-133. doi: 10.13832/j.jnpe.2019.03.0130

Abstract(628)

Abstract:
The control rod drive line in this paper has been adopted in the reliability growth test, which was simulated as 1:1. This test was carried out in the operating condition of the practical reactor. The service life and failure event of the control rod drive line was tested by the reliability growth test. From the failure event, we can find out the failure reasons. Furthermore we can do targeted improvement by the failure reasons. After that the experimental verification was conducted once again on the basis of the original prototype of the control rod drive line multiplexing. The result of experimental verification proved that from the reliability growth test of control rod drive line, we can find out the potential shortfalls. Furthermore we can do targeted improvement by the potential shortfalls to improve the service life of the control rod drive mechanism.

Optimization of CPR1000 Charging Pump Shaft Power and Hydraulic Performance Study

Zhang Qiangsheng, Ma Huiping, Wang Yan, Shen Wei, Wu Chao

2019, 40(3): 134-137. doi: 10.13832/j.jnpe.2019.03.0134

Abstract(744)

Abstract:
Based on the analysis of CPR1000 charging pump hydraulic performance test, it is found that the large inlet area of the diversion body is the main reason for the high shaft power. Every piece of steel plate was welded to the inlet tongue respectively to improve the hydraulic performance of the charging pump. Firstly, with the help of Ansys CFX, the flow field inside the charging pump of the modified scheme was numerically simulated and compared with the measured hydraulic performance data before the scheme modification. Then the hydraulic performance of the charging pump was tested and verified. The results show that the shaft power of the charging pump is controlled below 650 kW, meeting the requirements of the evaluation test program and technical specification, and the safety and reliability of the charging pump is improved.

Investigation on Experimental Method for Validating Performance of Passive Autocatalytic Hydrogen Recombiner

Li Zhiming, Wang Hongqing, Ma Weigang, Qiu Tian, Li Yang, Jiang E

2019, 40(3): 138-141. doi: 10.13832/j.jnpe.2019.03.0138

Abstract(754)

Abstract:
The startup and stop threshold of hydrogen recombination and the hydrogen removing rate are key performance indexes for the passive autocatalytic hydrogen recombiner (PAR). A straightforward and convenient experimental method for validating the performance of PAR is proposed. Put PAR in an airtight vessel, and hydrogen is feed to the vessel. If the PAR start hydrogen recombination and whole inlet hydrogen concentration variation curve versus time is under given line A, the startup threshold is confirmed to be not greater than A. If the final horizontal segment of the inlet hydrogen concentration variation curve versus time is under given line B, the stop threshold is confirmed to be not greater than B. If PAR realized steady operation status, the average flow rate of hydrogen fed into the vessel is the hydrogen removing rate. Test rig is designed and setup, and the experiment for validating the performance of PARQX-15 PAR is carried out. Successfully, the startup threshold is confirmed to be not greater than 2(v)%, and stop threshold is confirmed to be not greater than 0.5(v)%, and the hydrogen removing rate is greater than 536 g/h. Conclusion can be made that this method is practical and effective.

Engineering Development and Application of Hydrodynamic Mechanical Seal of Reactor Coolant Pump

Feng Xiaodong Ma Yu, Song Kuilong, Tan Heping, Li Mengqi, Lyu Yanguang

2019, 40(3): 142-145. doi: 10.13832/j.jnpe.2019.03.0142

Abstract(781)

Abstract:
The shaft seal of reactor coolant pump (RCP) consists of three same hydrodynamic mechanical seals in series. It is the heart of RCP and its leakage directly determines the normal operation of the RCP. This paper introduces a new extruding deformation lapping method to manufacture the hydrodynamic mechanical seal. The stationary ring is deformed by using the extruding deformation tool and metallic gasket, and nine waveform grooves are produced on the seal end face. The function test showed that the low-pressure (LP) leakage meets the design requirement of the shaft seal under the examination pressure. Meanwhile the impact examination under the saltation pressure showed that the liquid film between new hydrodynamic mechanical seal friction pair was intact, and no failure occurred on the seal. The operation of this hydrodynamic mechanical seal in the nuclear power plant shows that the operation performance agrees well with the function test results, and proves that this new hydrodynamic mechanical seal is with a high reliability in engineering application.

Development and Application of Fuzzy Multiple Model Simulation System for Nuclear Reactor Core

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

2019, 40(3): 146-149. doi: 10.13832/j.jnpe.2019.03.0146

Abstract(701)

Abstract:
Abstract: The reactor core model for a single power level can not accurately represent the dynamic characteristics of the core under different power conditions. In order to solve the problem, the triangle membership function is used to weigh five local models at different power levels, and the fuzzy multi-model simulation system is developed. Taking the PWR core as an example, the disturbance simulation of the reactivity and the coolant inlet temperature are carried out. The results show that the fuzzy multi-model simulation system can be applied to the simulation of different core power levels.

Development and Application of Eddy Current Inspection System for Heat-Exchanging Tubes of Horizontal Steam Generator

2019, 40(3): 150-154. doi: 10.13832/j.jnpe.2019.03.0150

Abstract(712)

Abstract:
It is difficult to conduct eddy current inspection on the horizontal steam generator for its deep collector, numerous heat-exchanging tubes with many elbows and small bending radius and materials with poor uniformity. A horizontal steam generator eddy current testing system(C-SMART) for heat-exchanging tubes was developed, including mechanical positioning drive device, control system and control software, and eddy current data acquisition and analysis software. The system is with the characteristics of fast and accurate positioning, single tube calibration, high efficiency and so on. The system has significant economic and social benefits.

Design and Verification of STP Feedthrough of EPA forThird Generation NPPs

Wang Xinyu, Zhou Tian, Wang Guangjin, Chen Qing, Zhou Yuan, Wang Jiangwu

2019, 40(3): 155-158. doi: 10.13832/j.jnpe.2019.03.0155

Abstract(568)

Abstract:
Based on the functional requirements of shielded twisted-pair(STP) cable penetrating containment, and combined with the environmental conditions of the third-generation NPPs, this paper developed a STP feedthrough for electrical penetration assembly(EPA), which is specially used for connecting STP cables inside and outside of the containment. This paper focuses on the overall structure design, material selection and type test of STP feedthrough. Through a series of type tests, it is proved that the STP feedthrough has good sealing performance, electrical performance, irradiation resistance and electromagnetic compatibility. The test results show that the STP assembly meets the expected design requirements of specified design life, sealing and electricity.

Simulation Analysis of Performance of Nuclear-Gas Combined Cycle Power Generation Systems

Li Bin, Ba Xingyuan, Zhang Shangbin, Xu Wentao, Liu Zhe, Teng Zhaoyu

2019, 40(3): 159-164. doi: 10.13832/j.jnpe.2019.03.0159

Abstract(650)

Abstract:
For the low cycle thermal efficiency of Pressurized Water Reactor(PWR) nuclear power units and the growing demand on the nuclear power peaking capacity of the power grid, using Ebsilon software, based on the thermodynamic system model of the secondary loop of Daya Bay Nuclear Power Station, the nuclear-gas combined cycle power generation thermodynamic system is established. Taking the gas turbine cycle efficiency and the combined cycle efficiency as the index of the heat economy, the performance of the combined cycle system is evaluated. At the same time, the effect of ambient temperature, pressure and gas turbine load on the system performance is analyzed. The results show that, the thermal efficiency of the nuclear-gas combined cycle system increased by 13.15% compared with the original nuclear power unit, the output power of the steam turbine increased by 75.49%, and the working environment of the steam turbine is improved obviously; The fall of ambient temperature will increase the efficiency and the power of the gas turbine, and the drop in ambient pressure will decrease the efficiency of the gas turbine and the combined cycle efficiency. When reducing the load of the gas turbine, the inlet temperature of nuclear reactor remains unchanged by afterburning the natural gas, and the steam turbine can still have high output power. The adjustable range of the load is from 56.57 % to 100%.

Design of an Emergency Power Supply System for a Research Reactor

Qin Fujun, Li Changshun, Zhang Ying, Jin Yang, Zheng Tingting

2019, 40(3): 165-169. doi: 10.13832/j.jnpe.2019.03.0165

Abstract(787)

Abstract:
In order to guarantee the safe operation of a new research reactor in China, the design criterion, system structure, function and equipment composition of Emergency Power System of the reactor are introduced, and the key elements are analyzed, which should be considered in the capacity determination process of diesel generator set, uninterrupted power supply(UPS) and Class 1E batteries. An Emergency Power System was designed as a special safety facility of the new research reactor, and then its capacity were calculated. The results show that the capacity of diesel generator set is 1000 kV?A, the maximum capacity of UPS is 600 kV?A, and the maximum capacity of Class 1E batteries is 5000 Ah. When the two outside powers are lost, this emergency power system can supply 72 h power to the reactor safety system continuously, and it can guarantee the safe operation of the reactor.

Research on Optimization of Parameter Uncertainty Applied in Small Range Power Uprating for Nuclear Power Plants

Yu Junhui, Guan Zhonghua, Huo Yujia, Zhu Jialiang, Duan Yongqiang

2019, 40(3): 170-174. doi: 10.13832/j.jnpe.2019.03.0170

Abstract(653)

Abstract:
The background and the method for the power uprating in nuclear power plants (NPP) in China were described, and the specifications on NPP power uprating in related laws and standards were analyzed and summarized. The classification of NPP power uprating was overviewed, and the specific requirements and main methods for NPP small range uprating were analyzed. Based on the analysis of NPP power and the calculation method for the uncertainty, the main parameters having effects on the calculation error for the thermal power of the core were specified, which were the main feedwater flow and the main feedwater temperature. The methods to apply the optimization of parameter uncertainty in the power uprating was analyzed and summarized, which was to add new differential pressure transmitter or ultrasonic flowmeter and temperature transmitter to optimize the parameter uncertainty. The economy for NPP small range power uprating was summarized.

Research of ECCS Scheme for Small Modular PWR Based on Combination of Deterministic and Probabilistic Methods

Gao Yingxian, Zhang Hang, Qiu Zhifang, Liu Zhaodong, Li Meifu, Zeng Wei

2019, 40(3): 175-179. doi: 10.13832/j.jnpe.2019.03.0175

Abstract(756)

Abstract:
In this study, based on the design characteristics of the small modular PWR, the emergency core cooling system (ECCS) scheme was analyzed for accident cases by applying the deterministic and probabilistic methods. The primary analysis shows that there is quite different requirement for the system scheme of the accumulator by deterministic and probabilistic methods. When the safety goal is ensured, it is suggested to eliminate the safety class accumulators (ACC) for ECCS simplification and other existing source water, such as IRWST in residual remove system, can be used after CMT failure.

Research of Visual Conversion Technology for Reactor Core Monte Carlo Model Based on CSG

Yuan Guanghui, Liu Dong, Yu Hongxing, Hao Jiangtao, Qiang Shenglong, Liu Ying, Cao Guohai

2019, 40(3): 180-184. doi: 10.13832/j.jnpe.2019.03.0180

Abstract(707)

Abstract:
In order to improve the efficiency of 3D visualization analysis of the core Monte Carlo fine transport calculation model (MC calculation model), on the basis of fully investigating the current development status of the MC calculation model 3D visualization tools, and by analyzing the characteristics of different types of geometric models and the description methods of constructive solid geometry (CSG) models, this paper proposed an automatic 3D visualization converting method for the core fine MC calculation model, and conducted a deep study on the key technologies such as the conversion from CSG model to boundary representation (BREP) model, the mesh discretization of BREP model and the optimization of conversion efficiency for CSG model. The test results show that the proposed method can meet the engineering requirements in terms of function and efficiency.

CFD Analysis for Two Phase Performance in 5×5 Rod Bundle with 7 Spacer Girds

Li Songwei, Li Zhongchun, Du Sijia, Zhang Hong, Liu Luguo, Shen Caifen

2019, 40(3): 185-190. doi: 10.13832/j.jnpe.2019.03.0185

Abstract(261)

Abstract:
A 5×5 rod bundles with 7 spacer girds which structure mixing gird (MG) and mid span mixing gird (MSMG) are arranged alternant are studied by two phase computational fluid mechanic(CFD), and bubbles coalescence and breakup, and heat transfer are considered in the calculation, but not the interface mass change. In order to choose the reasonable two phase model parameters, 5×5 rod bundles with 2 spacer girds (MG and MSMG) based on AFA3G fuel assembly isstudied by two phase CFD, and the sensitivity and uncertainty analysis is conducted on max bubble size, bubbles coalescence and breakup model parameters, non-drag force model, drag force model parameters, inlet bubble diameter and inlet void fraction distribution. Use this model setting, the study of the two phase performance is conducted on AFA3G fuel assembly with 7 spacer grids. The calculation shows that the bubble through the spacer grid is without periodicity, but the void fraction fields of every MSMG upstream are similar and the MG upstream shows the same situation. This research is useful to choose the model setting and geometry size for the rod bundle with spacer girds in the two phase calculation, for geometry size it is possible to analyze the two phase performance on rod bundle with 2 or 3 spacer grids. Finally, the void fraction distribution of AFA3G fuel assembly and advanced fuel assembly are compared, and evaluated in terms of the improvement of the CHF. It agrees well with the experiment, which verifies the evaluation method, the research is the base for the establishment of the evaluation guideline for the thermal hydraulic performance of the fuel assembly by two phase CFD calculation.

Design and Analysis of an Hourglass-Type Electromagnetic Vibration Isolator

Luo Yajun, Gao Xing, Zhang Yingqi, Zhang Xue, Zhang Yahong, Zhang Xinong

2019, 40(3): 191-197. doi: 10.13832/j.jnpe.2019.03.0191

Abstract(611)

Abstract:
In order to maintain a good isolation performance of the isolator under low amplitude disturbance. the design method of an hourglass-type electromagnetic isolator with high electromechanical coupling performance is proposed in this paper. Combining a suboptimal Bang-bang control algorithm, the theoretical modeling and performance simulation of the isolator are also completed. The results show that the displacement amplitude at the resonance peak can be decreased by 62.64% when the maximum active force is 1.0 N under the excitation of sinusoidal displacement, and the control effect becomes better with the increase of maximum active force. Also, the root of mean square(RMS) values of output displacement can be decreased by 49.12% and 69.29% when the maximum active forces are 0.5 N and 1.0 N under the random excitation, respectively. The presented electromagnetic isolator has good vibration isolation effect under the active closed-loop control. Compared with the traditional electromagnetic isolator, the isolator has the advantages of not needing additional spring and small control current.

Structural Analysis and Evaluation Method for Nuclear Class 1 Piping Systems at Elevated Temperature

Zhang Xiaochun, Gong Wei

2019, 40(3): 198-204. doi: 10.13832/j.jnpe.2019.03.0198

Abstract(799)

Abstract:
To solve the problems encountered in the design of the piping system at elevated temperature, the present paper demonstrates a methodology of structural analysis and safety evaluation procedures for high temperature piping. A bridge between piping analysis software and ASME-NH is established. Firstly, the detailed analytical solutions of stress components for piping components (straight pipe and bent pipe or elbow) under multiple loading combinations are proposed. Then, the analytical predictions of the stresses are compared with the solutions given by finite element analysis (FEA). The results show that the proposed analytical results agree with FEA results very well. And finally, the piping structural analysis and integrity assessment procedures are given by coupling the pipe line model and the newly proposed analytical solutions. And thus a method to implement the piping design criteria of ASME-NH-3600 for the analysis of nuclear class 1 piping components is provided.

Application of Modal Strain Energy in Dynamical Analysis of Reactor and Primary Loop System

Xiong Furui, Ye Xianhui

2019, 40(3): 205-210. doi: 10.13832/j.jnpe.2019.03.0205

Abstract(678)

Abstract:
In order to investigate the dynamical contribution of individual component and equipment in system-level models, this study proposed a computational method based on the modal strain energy (MSE). Two engineering case studies applying the MSE method are demonstrated herein. The first study investigates the main reason causing large seismic response change of a prototype reactor coolant system (RCS) surge pipe due to the location change of surge suspension device. The analysis shows that the local dominant mode of the surge pipe is altered by the location change of the suspension. Such change of dominant resonant frequency leads to a significant change at the corresponding frequency of input seismic response spectrum, thereafter affects the dynamical response. The second case study discusses the decoupled strategy of the main steam-pipe from the loop model for separated dynamical analyses. MSE analysis, together with the second decoupling criteria suggested by USNRC SRP 3.7.2, justifies this deed. Modal strain energy (MSE) method proposed in this paper is able to quantify the dynamical contribution of components and equipment from the system-level dynamical analysis models. The computation of MSE only utilizes the system-level model stiffness and mass matrices. No time-domain transient analysis is necessary.

2019 Vol. 40, No. 3