核动力工程

CFD Investigation on Flow and Heat Transfer Characteristics of Fuel Assembly for VVER Reactor

Wang Xiong, Du Daiquan, Zeng Xiaokang, Yang Xiaoqiang, Zan Yuanfeng

2018, 39(3): 6-9. doi: 10.13832/j.jnpe.2018.03.0006

Abstract(1396) PDF(569)

Abstract:
The flow and heat transfer characteristics of AFA fuel assembly for VVER reactors have been investigated using computational fluid dynamics (CFD) simulation. The flow field, pressure drop and temperature distribution of the coolant in AFA under normal regime have been calculated. The results show that the pressure drop of the spacer grid of AFA is lower than that of the grid having mixing vane. The stagnation zone of coolant appears around the rim of the spacer grid and causes higher temperature in the periphery region of AFA. The power ratio of the circumferential pin around instrumental tube (K_c) with different values has a great effect on the measured temperature of the coolant at FA outlet. The results can be referred in the setting of temperature warning value (ΔT_t) for the reactor core during the operation of nuclear power plants.

Diagnosis of Eigenvalue of Two Static Laboratory Devices

Bai Yun, Li Zhongbao, Fan Xiaoqiang, Gong Jian, Peng Xianjue

2018, 39(3): 10-12. doi: 10.13832/j.jnpe.2018.03.0010

Abstract(291) PDF(156)

Abstract:
It is founded that the prompt neutron decay characteristics of the metal reflection layer subcritical device and prompt neutron decay characteristics of the hydrogen reflection subcritical device are not the same. For the fast spectrum device, after the DPF neutron source is induced into the device, the energy spectrum of the device will be basically unchanged, the time constant of the eigenvalue calculation concept and numerical eigenvalue can be applied to research the fast spectrum device, and the numerical calculation of time constant eigenvalues will be same with the experimental measurements for comparison. However, for the hydrogen reflective layer device, the energy spectrum will changed till reaching the steady state of the device, based on the definition of eigenvalue, the energy spectrum of the hydrogen reflector device reached steady state within 10 milliseconds or longer, the longer the more information can slow neutron the spectrum, so the eigenvalue is not applicable for hydrogen reflector device.

Experimental Study on Mixed Convection Heat Transfer in Rod Bundles

Liu Da, Gu Hanyang

2018, 39(3): 13-17. doi: 10.13832/j.jnpe.2018.03.0013

Abstract(249) PDF(186)

Abstract:
With the background of the low flow rate heat transfer in case of the natural circulation in the reactor, the mixed convection was experimentally studied in a 5×5 rod bundles. The experiment was performed at 6 MPa, the mass flow rate from 25 to 150 kg/(m²·s), the heat flux from 25 to 300 kW/m², Reynolds number from 1000 to 30000, Buoyancy parameter from 2×10^-7 to 3×10^-3. It is concluded that with the increasing of the Buoyancy parameter, the heat transfer in rod bundles was impaired firstly and then followed by enhancement. It is observed that in the rod bundle channel, the influence of the buoyancy force occurs at Bo*=3.5×10^-6. Even when Reynolds number is higher than 15000, the heat transfer could be impaired by the buoyancy force. An empirical correlation used to calculate mixed convection heat transfer in rod bundles was proposed based on the experimental data.

Periodically Reloading and Transmutation Characteristics Numerical Analysis of TRU Fueled Thermal Molten Salt Reactor

Yu Tao, Xie Jinsen, Zhao Wenbo, Chen Zhenping, Xie Qin, Liu Zijing

2018, 39(3): 18-22. doi: 10.13832/j.jnpe.2018.03.0018

Abstract(306) PDF(147)

Abstract:
The molten salt reactor (MSR) features convenient fuel fabrication, good neutron economy and flexible fuel management, has and there is the potential to utilize the TRU produced in the spent fuel of the light water reactor directly. The selection of fuel, lattice parameters and fuel/graphite volume fraction has been optimized in this paper. Through the calculating and analyzing the core lifetime of the TRU fueled thermal spectrum MSR, the stock of TRU, MA-fission ratio and TRU-fission ratio, it proves that thermal spectrum MSR can operate for a long period with long refueling cycle, which can reduce the difficulty of online refueling and have a better transmutation ratio of MA and TRU with lower radioactive toxicity.

Uncertainty Study of Core Power Distribution for Software NESTOR

Liao Hongkuan, Li Qing, Yu Yingrui, Hu Yuying, Ning Zhonghao, Lu di, Huang Shien, Zhao Wenbo

2018, 39(3): 23-28. doi: 10.13832/j.jnpe.2018.03.0023

Abstract(484) PDF(181)

Abstract:
As the key indicator of the nuclear design, the computational accuracy of the core power distribution is very important for the evaluation of the economy and safety of nuclear power plants. As the first nuclear power software developed on self-reliance in China, the computational accuracy and applicability of NESTOR is the foundation for its application. Based on the random sampling statistical analysis (RSSA) method and deviation transmission idea, the uncertainty of core power distribution was obtained by combining two independent uncertainties resulting from the analysis of the uncertainty of physical model and the uncertainty of the change of parameters. The results indicate that the RSSA is feasible in the uncertainty analysis of nuclear design. In addition, in the analysis of the uncertainty of physical model, the core power distribution was decomposed into 2 parts, including the uncertainty of detailed power distribution in an assembly and the uncertainty of assembly power. As a result, the uncertainty of radial power distribution caused by physical model was ±3.653%, the uncertainty of radial power distribution caused by the change of parameters was ±0.964%, and the final uncertainty of radial power distribution was ±3.778% under the condition of 95% confidence coefficient and 95% probability that computed through the deviation transmission idea. The computational accuracy is as high as the engineering design software, and it lays foundation for the application and verification of NESTOR.

Improvement and Validation of Free Convection Heat Transfer Model for Tube Bundle in RELAP5

Xian Lin, Jiang Guangming, Li Jie, Wang Xiaoji, Yang Fan, Li Feng, Ran Xu

2018, 39(3): 29-32. doi: 10.13832/j.jnpe.2018.03.0029

Abstract(491) PDF(243)

Abstract:
Based on the validated model of single phase and two phase free convection heat transfer from tube bundle, RELAP5 is improved with the ability to simulate the single phase and two phase free convection heat transfer for tube bundle. The test simulation result analyzed with the improved RELAP5 was compared with that from the original RELAP5. The comparison shows good agreements between the experiment data with the simulation result from improved RELAP5.

Investigation on Nonuniformity of Heat Transfer in Triangular Subchannels of Supercritical Water Cooled Reactor

Wang Weishu, Hou Yanliang, Xu Weihui, Zhu Xiaojing, Bi Qincheng

2018, 39(3): 33-39. doi: 10.13832/j.jnpe.2018.03.0033

Abstract(271) PDF(167)

Abstract:
A physical model was built to numerically investigate the nonuniform heat transfer characteristics of supercritical water within a triangle sub-channel. The corresponding fuel rod diameter was 8 mm and the pitch-to-diameter ratio was 1.4, respectively. Experimental result shows that the axial heat transfer is uneven and the axial heat transfer intensity has a peak value. The heat transfer coefficient is big when the supercritical water is in the pseudo-critical region and is small when the water is far away from the pseudo-critical region. The comparison between the experimental data and numerical result show that the SSG turbulence model can well predict the heat transfer characteristics of supercritical water in a triangle sub-channel of SCWR. The nonuniform heat transfer of supercritical water is seriously in the sub-channel. In the mainstream direction, the heat transfer from the inner wall to the fluid is enhanced in the great specific heat region and weakened in the high enthalpy region. The local heat transfer coefficient increases first and then decreases with the increasing of the circumferential angle, which reaches a peak in the central region and a trough near the narrow gap. The degree of circumferential heat transfer nonuniformity also differs in different enthalpy regions. Moreover, with the increasing of the pitch-to-diameter ratio, the average wall temperature rises and the heat transfer coefficient decreases, and the circumferential heat transfer nonuniformity reduces significantly.

Dynamic Analysis of Fuel Assembly for Accident Condition Based on ANSYS

Qi Huanhuan, Wu Wanjun, Shen Pingchuan, Jiang Naibin, Ye Xianhui, Huang Xuan, Huang Qian

2018, 39(3): 40-44. doi: 10.13832/j.jnpe.2018.03.0040

Abstract(1012) PDF(234)

Abstract:
The process of fuel assembly dynamic analysis for accident condition was studied. Method of axial and lateral dynamic modeling was developed. The axial and lateral dynamic response calculation method was established, and then the grid impact force and guide thimble stress calculation were carried out. Based on ANSYS APDL and UIDL language, introducing the idea of parameterization and modularization, the fuel assembly dynamic analysis program (program-developed) for accident condition was developed. Compared validation was carried out using program-developed and software-specific respectively for a certain type of fuel assembly. Comparison results show that the difference was small, and in the range of engineering permissible error. The program developed can be used to analyze the fuel assembly accident instead of software-specific. Analysis ability of the program developed was stronger and calculation efficiency was higher than that of the specific software. Selecting a nuclear power plant as the analysis object, the program developed is used to do the dynamic calculation of fuel assembly for accident condition, and the analysis results meet the requirements of the code.

A Papid Method for Detecting Damaged Fuel in Research Reactors

Fan Yijun, Jia Haopeng, Tang Yang, Wang Xiaobing, Li Chengye

2018, 39(3): 45-47. doi: 10.13832/j.jnpe.2018.03.0045

Abstract(739) PDF(187)

Abstract:
The spent fuel element in the research reactor needs to be tested before it was safely transferred to the spent fuel storage tank. The current testing method is time-consuming and difficult to judge the specific damaged object quickly. In this paper, a rapid method for detecting damaged fuel elements is proposed, which can realize the screening of spent fuel in advance and improve the throughput rate of spent fuel.

Reserach of Bonded Technology of Strain Gage for Fuel Assembly Model Guide Tube

Tang Li, Wang Jun, Ma Wenhui

2018, 39(3): 48-50. doi: 10.13832/j.jnpe.2018.03.0048

Abstract(173) PDF(162)

Abstract:
The mechanical test for fuel assembly needs to measure the physical parameters of the stress point in different parts of fuel assembly. The strain test for these stress points only can be realized by sticking the strain gauges. However, the structure of the fuel assembly is complex and compact, and it is difficult to implement the stickup of the strain gauges. In this article, a kind of new strain gauge stickup technique is developed to make the stain gauges easily and reliably stuck on the narrow space of the fuel assembly guide tube. The test results show that the test data of the strain gauge are effective and reliable.

Determination of Fissile Nuclide ²³⁵U Content in Re-Irradiated Spent Fuel Assemble with Nondestructive Assay

Dou Haifeng, Li Rundong, Zhu Shilei, Wang Junwei, Si Kaituo, Yuan Shu, Yang Xin, Leng Jun

2018, 39(3): 51-55. doi: 10.13832/j.jnpe.2018.03.0051

Abstract(859) PDF(162)

Abstract:
Adequate knowledge of burnup levels of fuel elements within a research reactor is of great importance for its safe operation. The traditional nondestructive assay of burnup is to measure the radiation emitted either as neutrons or gamma rays. But the results are not satisfactory in accuracy because of variability of core loading and operation history. This paper presents a method for the experimental determination of fissile nuclide ²³⁵U content in Spent Fuel Assembles (SFAs). The method is based on re-irradiation of SFAs and measurement of the delayed gamma-rays emitted by the generated fission products. The most important advantage of this method is its independence of SFA irradiation history. This paper emphasizes how to discriminate the resource of characteristic gamma ray and introduces the experimental device. A SFA with about 15% burnup unloading from CMRR is measured by the above method and the uncertainty is less than 5%.

Denoising Method and Prediction Technology of Nuclear Island Foundation Settlement

Wang Mingyu, Sun Hao, Liu Rui

2018, 39(3): 56-61. doi: 10.13832/j.jnpe.2018.03.0056

Abstract(214) PDF(152)

Abstract:
The foundation deformation of nuclear power plant nuclear island is observed during the whole construction process. Different sites are with different values of foundation deformation that are with different effects on the items crossing buildings at the completion of the raft foundation of the nuclear island. In this paper, the data from 15 times of observation of the foundation settlement in a nuclear power plant are analyzed. Firstly, the observation data is disposed by the wavelet denoising technology. The pretreatment result shows that the foundation settlement is highly time dependent and tends to steady according to the foundation settlement velocity. . Based on the above, a 3 times short-term prospect and a 30 years long-term prospect are predicted with hyperbolic method and gray theory method. In short-term prospect aspect, two methods, especially the gray theory method, are both conservative comparing with the actual observation result. As for the long-term prospect, the gray theory method is inapplicability for its result diverges. While the prospect with hyperbolic method is effective comparing with the foundation settlement data of two typical nuclear power plants which have operated for 30 years.

Study on Application of Fatigue Strength Reduction Factor and Stress Concentration Factor in Fatigue Analysis of Screw Threads

Chen Tao, Liu Pan, Xu Xiao

2018, 39(3): 62-66. doi: 10.13832/j.jnpe.2018.03.0062

Abstract(1270) PDF(227)

Abstract:
Taking the reactor pressure vessel as an example, the theoretical relationship of fatigue strength reduction factor (K_f) and stress concentration factor( K_t ), the reasonable values of K_f and K_t, the principles of equivalency of K_f and K_t in engineering project and the methods and techniques of how to use K_f and K_t in fatigue analysis are presented. An example of fatigue analysis for screw threads of reactor pressure vessel (RPV) stud with certain K_f is given. Consequently, definitions and differences of K_fand K_t are classified. Opinions on the application of K_f to high strength alloy stud are given, which can be used as a reference to the stud selection and the structure design in the engineering application.

Analysis for Fluid Pressure Fluctuation of Core Barrel Surface Based on Experimental Data

Wang Dasheng, Duan Yuangang, Liu Pan, Kong Xiaofei

2018, 39(3): 67-70. doi: 10.13832/j.jnpe.2018.03.0067

Abstract(209) PDF(133)

Abstract:
Based on the fluctuating pressure data obtained in the flow- induced vibration test of reactor pressure vessel internals scale model, the distribution characteristics of fluctuating pressure power spectral density are analyzed at different positions on the core barrel surface, and the correlation length characteristics of fluctuating pressure power spectral density are obtained based on the correlation analysis. The power spectrum density of the fluctuating pressure on the core barrel surface is broadband attenuation spectrum with a very rich frequency andecreases quickly with the increasing of the frequency and then tends to be gentle; The fluctuating pressure power spectral density is basically the same in the region located at the same height on the core barrel, but the differences of the PSD (power spectral density) are obvious between the regions at different heights. The correlation length of the fluctuating pressure power spectrum density decreases sharply with the increasing frequency and then tends to be constant. The flow-induced vibration response of the core barrel has little effect on the fluctuating pressure power spectral density and it is reasonable to simplify the flow-induced vibration of the core barrel into the weak coupling problem.

Numerical Simulation of Nuclear Secondary Bellows Globe Valve under Transient Thermal Shock

Li Shuxun, Luo Xiangyao, Lyu Xing, Zhang Lifang, Xu Xiaogang

2018, 39(3): 71-77. doi: 10.13832/j.jnpe.2018.03.0071

Abstract(889) PDF(170)

Abstract:
Abstract: In order to study the effect of the transient thermal shock on the structural strength and fatigue life of the nuclear secondary bellows globe valve, the hot-fluid-solid coupling analysis of the nuclear secondary globe valve body was carried out by Fluent and ANSYS software, based on the theory of fluid-solid coupling and thermal boundary condition. The effects of the temperature field, thermal stress and fatigue life of the monitoring point at different time points and the effect of the thermal shock time on the fatigue life of the degree of sensitivity were studied. The results showed that the effects of the transient pressure on the temperature field, structural strength, fatigue life and sensitivity of the valve body was huge, and have to be eliminated to ensure the high safety of the nuclear secondary bellows globe valve.

Research on Multiple Source Data Fusion Method

Li Hongwei, Liu Zhaodong, Min Yuansheng, He Liang, Liu Liu, Zhao Wentao

2018, 39(3): 77-80. doi: 10.13832/j.jnpe.2018.03.0001

Abstract(1056) PDF(368)

Abstract:
This paper selects the data fusion method which is easy to realize in the engineering application, and takes the deviation of the data and the real value of the data as the stability criterion of the data fusion method, and carries out the simulation test of the data fusion of the three redundancy and multiple source data. The test results show that the weighted least square method is used in the data fusion. The performance is stable, which can replace the commonly used moment estimation fusion and median fusion, and improve the accuracy of the fused data.

Research on Digital Upgrade Verification Platform Scheme for Simulation Control System of Daya Bay Nuclear Power Station

Fang Yu, Xiong Guohua, Ma Shu, Cai Yefa, Kuang Dejun, Peng Chao

2018, 39(3): 81-85. doi: 10.13832/j.jnpe.2018.03.0081

Abstract(281) PDF(174)

Abstract:
According to the scope and actual situation of the existing simulation control system in Daya Bay Nuclear Power Station, the signal configuration, interface type and typical control loop are analyzed. In this paper, the technical requirements of the verification platform are put forward, and the verification platform based on DCS system, process simulation system and KRG system is set up. The specific verification contents are also given in detail. Using the verification platform to carry out the corresponding verification work, the differences before and after the improvement of the platform can be effectively identified to ensure the smooth implementation of digital upgrades.

Research on Periodical T2 Test for Reactor Protection System

Zhang Liangliang, Zhang Yu, Zhou Can, Chen Jie, Liu Dongbo

2018, 39(3): 86-89. doi: 10.13832/j.jnpe.2018.03.0086

Abstract(1202) PDF(229)

Abstract:
T2 test of the safety DCS of the nuclear power plant is important for the reliability of logic function. Based on safety-related regulations and standards, this paper introduces different safety DCS platforms, and analyzes their advantages and disadvantages respectively. Finally, the factors that should be considered in the design of T2 test are proposed. A proposal of T2 test on a regular basis and suggestions for its improvement is given.

Nuclear Power Plant Pressurizer Pressure Control Based on Endocrine Fractional Order PID Controller

Qian Hong, Zheng Zibin, Zheng Miao

2018, 39(3): 90-94. doi: 10.13832/j.jnpe.2018.03.0090

Abstract(292) PDF(147)

Abstract:
Based on the principles of biological neuroendocrine gland hormone regulation and the fractional order calculus, and considering that the pressurizer system of nuclear power plant is a very complicated system, a endocrine fractional order PI^λD^μ controller was designed with long feedback and ultra-short feedback, in order to obtain satisfactory control results. The primary controller is proportional control. According to the system errors, it can adjust the control parameters of the secondary one dynamically and eliminate the control error quickly and stably. The secondary is fractional order PI^λD^μ controller and a indirect algorithm (Oustaloup algorithm) was used to implement the fractional order controller. The system can be more excellent in velocity, precision and anti-interference ability. Simulation results show that the endocrine fractional order PI^λD^μ controller is with better performance and strong anti-interference ability than the traditional PID control.

Common Cause Failure of Digital Safety Level DCS Emergency Shutdown System

Ma Quan, Luo Qi, Song Xiaoming, Liu Yanyang

2018, 39(3): 95-99. doi: 10.13832/j.jnpe.2018.03.0095

Abstract(353) PDF(182)

Abstract:
This paper takes the digital safety level DCS emergency shutdown system which used 2-out-of-3 architecture as the research object, and establishes the reliability model of the system by the method of Markov. The average probability of failure on demand, as so called the PFD_avg, under two cases of common cause failure and non common cause failure consideration are calculated and compared. In addition, it turns out that the PFDavg changes to be bigger with the increasing of the factor of common cause failure. Thus, in order to decrease the factor of common cause failure, it is necessary to control the common cause failure by some effective measures when designing the system to improve the reliability of RTS.

Study on Key Techniques and Framework of Probabilistic Safety Assessment in Xi'an Pulsed Reactor

Wang Baosheng, Tang Xiuhuan, Shen Zhiyuan, Zhu Lei

2018, 39(3): 100-106. doi: 10.13832/j.jnpe.2018.03.0100

Abstract(713) PDF(152)

Abstract:
Based on the unique design characteristics and safety features of Xi'an Pulsed Reactor (XAPR), the special key techniques of XAPR probabilistic safety assessment (PSA) was studied. XAPR PSA analysis framework and the way to implement the key technique were put forward. Finally, XAPR PSA research thought was illustrated by a case study on medium LOCA at the core height of XAPR reactor pool. The analysis shows that the integral event tree framework, which spread through event sequence structure with initiating events at the beginning and radioactive release categories in the end, is suited to XAPR PSA analysis at this stage.

Development of Key Equipment for In-Plant Spent Fuel Unloading from Cask

Weng Songfeng, Ren He, Dong Dailin, Luo Ying, Yang Qihui, An Yanbo, Zhang Chao

2018, 39(3): 106-109. doi: 10.13832/j.jnpe.2018.03.0106

Abstract(929) PDF(227)

Abstract:
Spent fuel in-plant transport is a valid solution to solve space shortage for spent fuel storage in nuclear power plants. This paper analyzes the design criteria and the safety risk of spent fuel in-plant transport, and describes the working principle and application of the damaged fuel detecting equipment and the spent fuel cooling equipment. The application results show that the damaged fuel detecting equipment can identify the damaged fuel in the spent fuel cask efficiently and the spent fuel cooling equipment can cool down the spent fuel cask with spent fuels.

Design and Verification of Acoustic Leak Monitoring System for Nuclear Power Plants

Zhou Zhengping

2018, 39(3): 110-113. doi: 10.13832/j.jnpe.2018.03.0110

Abstract(1011) PDF(429)

Abstract:
This paper considers the design basis and function of leakage monitoring system in nuclear power plants with VVER-1000. The process of the leak judgment and algorithm of leakage flowrate and leakage location are introduced. The scheme of acoustic model chart of the reactor coolant circuit is established and background noise in the reactor coolant circuit loop is obtained from calculation, and compared with the actual test results. The test bench of pipeline model is established and pipeline model is verified by the test bench. The correlation coefficient used to calculate the leakage flowrate and leakage location is obtained based on the test result. Through the design and verification of the acoustic leak monitoring system for nuclear power plants, the solid foundation is laid for the development of the online acoustic leak monitoring system for Unit 1 and Unit 2 in Tianwan Nuclear Power Plant.

Quantitative Analysis on Uniformity of Inflow of Fan Coil Unit in Containment Cooling System

Cui Guoqiang, Zhang Li, Xiao Bole, Liu Jiang, Zhang Qiang

2018, 39(3): 114-118. doi: 10.13832/j.jnpe.2018.03.0114

Abstract(673) PDF(208)

Abstract:
The uniformity of the inflow of the fan coil unit in the containment cooling system was studied quantitatively by numerical simulation and experimental measurement in this paper. The flow field in the fan-coil unit was analyzed by numerical calculation. The inlet velocity data of 50 points on the windward face of the fan-coil unit were obtained by experimental measurement. The data from the measurement and the calculation were processed by using the definition of “Face Velocity Uniformity” and “Relative Standard Deviation”. The processing results quantitatively evaluated the uniformity of the inflow of the fan coil unit. The results of “Face Velocity Uniformity” did not meet the requirements in GB/T 14294—2008 for not less than 80%. The results of “Relative Standard Deviation” exceeded the requirement that the value was usually less than 15%.

Development of Performance Test System for MSIV Solenoid Valves in Nuclear Power Plants

Shu Zhifeng, Huang Ping, Zhu CuiYun, Yang JinRui

2018, 39(3): 119-121. doi: 10.13832/j.jnpe.2018.03.0119

Abstract(777) PDF(242)

Abstract:
In order to meet the need of MSIV solenoid valves performance test in PWR nuclear power plants (NPPs), the integrated performance test system for solenoid valve opening-closing characteristics and sealing performance is developed. This paper introduces the principle and composition of the system. The system can effectively detect the overall performance of MSIV solenoid valves, and then guide the follow-up maintenance work, effectively avoiding human error, and can be used as an important maintenance tool for MSIV solenoid valves.

Feasibility Analysis and Optimization for Turbo-Generator Rushing with Non-Nuclear Steam in Fuqing Nuclear Power Plant

Xiao Bo, He Liu

2018, 39(3): 122-127. doi: 10.13832/j.jnpe.2018.03.0122

Abstract(194) PDF(224)

Abstract:
Compared with the rushing with nuclear steam, the turbo-generator rushing with non-nuclear steam can be used to verify the design, manufacture and installation quality of the turbo-generator, and shortens the timescale of the system commissioning of the nuclear power plant to create huge economic benefits. By establishing the thermal equilibrium calculation equation based on the first law of thermodynamics, this paper calculates the time of turbo-generator holding at 1500rpm, the consumed energy of the rushing, the replenishment of loop 1 and loop 2.The theoretical calculation is further checked according to the actual process of Unit 1 in Fuqing No.1 non-nuclear steam rushing. It is proved that the theoretical calculation method agrees well with the actual process of the system. In the process of non-nuclear rushing of Unit 3 in Fuqing Nuclear Power Plant, the non-nuclear rushing step is optimized to reduce the fluctuation of the key parameters of loop 1 and loop 2 during the rushing process for reducing the risk of unit control and extend the non-nuclear rushing time for further verifying the quality of the steam turbine . After the non-nuclear rushing of Unit 3 in Fuqing Nuclear Power Plant, the optimization measures were proved to be effective by comparing the non-nuclear rushing parameters between Unit 1 and Unit 3.

Application of GMAW Automatic Welding Process in Nuclear Power Plants with Steel Containment Vessel

Liu Fei, Tang Shi

2018, 39(3): 128-133. doi: 10.13832/j.jnpe.2018.03.0128

Abstract(291) PDF(160)

Abstract:
This paper briefly introduces the statusof the application of GMAW process, and the requirements of national nuclear safety regulation for process applications. This paper analyzes the difficulties of GMAW welding process application from the aspects of the support from the design organizations for steel containment vessel, the determination of welding groove form and size, welding parameters matching, groove on the back and clear root and other automatic GMAW welding process. This paper introduces the preparation before technology application from the aspects such as welding methods, equipment, material selection, welding process matching test and welding procedure qualification. Finally, this paper introduces the automatic GMAW welding application requirements and implementation results. The application results show that the GMAW welding process in the plant steel containment vessel welding is feasible.

Analysis and Research on Rotor Dynamic Characteristics of Nuclear Power Plant Reactor Coolant Pump under Reverse Flow Condition

Chen Xingjiang, Cong Guohui

2018, 39(3): 134-137. doi: 10.13832/j.jnpe.2018.03.0134

Abstract(197) PDF(145)

Abstract:
In order to prevent the RCP pump reverse rotation under the impact load of the reverse flow, the pawl type anti-reverse device consisted of return springs and hydraulic shock absorbers is mounted on the motor. According to the structure and working principle of anti-reverse device, the theoretical model of the anti-reverse device and the kinematic equation of the rotor under reverse flow condition are established. The dynamic characteristics of RCP pump rotor under reverse flow condition are analyzed, and the movement trajectory of the rotor is obtained. The results show that because the impact load under reverse flow condition is less than the design load of the anti-reverse device, the RCP pump rotor is subjected to reciprocating motion with six motion states, and the rotational speed of the RCP pump rotor is gradually reduced until it stops. The anti-reverse function is obtained by the anti-reverse device.

Analysis of Development Difficulty for Main Steam Isolation Valves

of Nuclear Power Plants

2018, 39(3): 138-142. doi: 10.13832/j.jnpe.2018.03.0138

Abstract(274) PDF(183)

Abstract:

Study on Quantitative Evaluation of Nuclear Facility Decommissioning Program Based on Analytic Hierarchy Process

Zhang Yongling, Zhao Wan, Zhang Hangzhou, Zhang Kaiyun

2018, 39(3): 143-146. doi: 10.13832/j.jnpe.2018.03.0143

Abstract(181) PDF(146)

Abstract:
An evaluation framework was developed using analytic hierarchy process to solve the evaluation problem in the nuclear facility decommissioning. This framework is consisted of 20 indicators derived from 6 criterions, which could holistically and structurally evaluate the decommissioning of nuclear submarines. The case study conducted onone decommissioning project demonstrated the effectiveness of this evaluation framework, which can be used as a reference for the future decommissioning evaluation.

Study on Load Limit of Gripping Components in Irradiation Specimen Handling Tool for Nuclear Power Plants

Liu Huifang, Yuan Zhanhang

2018, 39(3): 147-150. doi: 10.13832/j.jnpe.2018.03.0147

Abstract(709) PDF(151)

Abstract:
The extraction force of irradiation supervision capsules is changed greatly due to the mounting pattern on the lower reactor internals. The gripping components at the end of the critical load path are limited by the size of the top plug of irradiation supervision capsules and become the weakest part of the tool. In order to ensure the safe operation, this paper analyzed the load limit of the gripping components, and obtains its maximum load limit. Based on the calculation result and load test, optimization measures to improve load limit of gripper are analyzed, which provides a reference for the design and improvement of similar equipment.

Study on Pump Induced Vibration Acoustics of Reactors

Feng Zhipeng, Wu Wanjun, Xiong Furui, Zhang Wenzheng, Lyu Xi, Song Haiyang, Wang Bihao

2018, 39(3): 151-155. doi: 10.13832/j.jnpe.2018.03.0151

Abstract(681) PDF(180)

Abstract:
Advanced analysis methods for prediction and reduction of vibration and noise are of great significance for reducing vibration noise level of the structure. In order to make accurate noise prediction, it is necessary to adopt the appropriate method and establish the model of noise prediction. Firstly, the variable parameters and their influence law of vibration noise prediction and analysis method, for reactor and the primary loop system, are studied. Comparative studies are made on the acoustic analysis methods including direct boundary element method, acoustic finite element automatic matching layer method, acoustic finite element method adaptive order, and acoustic finite element automatic matching layer method combined acoustic finite element method adaptive order. Secondly, the effect of pressure hull on the vibration noise is studied by the finite element method and the acoustic boundary element method. Meanwhile, the effect of reinforcing rib and double shell on the system vibration characteristics and radiated noise is obtained. Based on these studies, a vibration-acoustic model is established. Then, the research of vibration reduction evaluation, acoustic radiation analysis, vibration and noise reduction are carried out. The vibration level difference of typical transfer path, the sound pressure level and the external acoustic field radiated by the pressure hull, and the effect of vibration and noise reduction with different vibration reduction measures are obtained.

Safety Analysis for Reactor Scram Subsystem Based on Multiple Methods

Liu Hua, Han Wenxing, Yang Xiaohua, Chen Zhi, Liu Zhaohui

2018, 39(3): 156-161. doi: 10.13832/j.jnpe.2018.03.0156

Abstract(510) PDF(149)

Abstract:
For the reactor scram subsystem, the failure and fault coverage statistics form for the instrument control system design phase is deduced by the combined use of three independent basic analysis methods FMEA, FTA, and STPA. STPA method can effectively make up for the inadequacy of FMEA and FTA method. At the same time, in the instrument control system design phase, STPA method is very suitable for finding the fault and safety issues in software, system interaction and communication for the reactor scram subsystems.

Study on Al₂O₃ Nanofluid Critical Heat Flux Mechanism Model: Physics Models

He Xiaoqiang, Yu Hongxing, Jiang Guangming

2018, 39(3): 162-165. doi: 10.13832/j.jnpe.2018.03.0162

Abstract(220) PDF(131)

Abstract:
In this study, to overcome the shortcomings of the present models, based on the analysis of the force balance of the bubble, considering the contact angle and capillary wicking effects, a Al₂O₃ nanofluid critical heat flux (CHF) mechanism model is developed. It is shown that, this model can simulate the effect of nanofluid concentration (c_NF) on CHF, that is, as c_NF increases, CHF increases at the beginning, and when c_NF is greater than a certain value, CHF no longer increases and maintains a constant value, and this model can explain that the diameter of nanoparticle (d₀) has no effect on CHF, which are in a good agreement with experimental results. As the contact angle or inclination angle increases, calculated CHF by this model decreases.

Validation of Neutronic Code SARCS-4 by Mock-up Critical Physics Experiments

Li Mancang, Chen Zhang, Yao Dong, Wei Yanqin, Wu Wenbin, Zhao Wenbo, Huang Shien, Ni Dongyang, Ju Haitao, Zheng Hongtao, Qin Dong, Zhang Zhizhu, Wang Liangzi, Wu Lei

2018, 39(3): 166-170. doi: 10.13832/j.jnpe.2018.03.0166

Abstract(281) PDF(161)

Abstract:
An advanced neutronic code system, SARCS-4, was developed in Nuclear Power Institute of China. To validate SARCS-4, an experimental program has been carried out. Based on the principles of single factor and variety of critical physics experiment, 3 types of mockup core layout were designed. The effects of control rod, burnable poison rod, shroud and core layout were studied. The calculation and analyses have shown that the core leakage, shroud effect, control rod and burnable poison rod effect are the key factors influencing the calculation precision. The accuracy of SARCS-4 has been confirmed by the mock-up critical physics experiments. Extended experimental program and further validation and improvement of SARCS-4 will be carried out in the future.

Software Development of Loose Parts Monitoring System Based on Localized PXI Control Modules

Li Xiang, Jian Jie, Li Hai, Wang Lei

2018, 39(3): 171-175. doi: 10.13832/j.jnpe.2018.03.0171

Abstract(776) PDF(181)

Abstract:
Using the loose part monitoring system (LPMS) developed by NPIC, the development of 16 channel LPMS software is carried out based on localized PXI (PCI extensions for instrumentation) modules. This paper introduces the software design requirements, design principles, design process and the design of the main interface, and the software implementation of the interface program of localized PXI control module is described in detail. The successful development of LPMS software based on localized PXI modules meets the design requirements, and it has been successfully applied in LPMS of a nuclear power plant abroad, which has played an active role in ensuring the safe and economical operation of the nuclear power plant.

Development of Fault Detection Instrument for Control Rod Drive Mechanism of Nuclear Reactors

He Pan, Peng Cuiyun, Zeng Jie

2018, 39(3): 176-180. doi: 10.13832/j.jnpe.2018.03.0176

Abstract(794) PDF(176)

Abstract:
Taking the CRDM as the study object, a fault detection instrument based on the detection theory of the mechanism structure noise is developed and tested on an experimental CRDM. Test result shows that the fault detection instrument can distinguish the fault of CRDM. An scientific effective fault detection method is provided for the installation and overhaul of CRDM.

Study on Neutron Noise Characteristics of PWR Nuclear Power Plants Based on Power Density Spectrum

Yang Taibo, Liu Caixue, Luo Ting, Hu Jianrong, Jian Jie

2018, 39(3): 181-183. doi: 10.13832/j.jnpe.2018.03.0181

Abstract(240) PDF(154)

Abstract:
The calculation method of neutron noise power density spectrum for PWR nuclear power plants is analyzed. By using this method, the power density spectrum of noise neutron is calculated on the basis of long-term internal vibration monitoring system in nuclear power plants. Therefore, the power density spectrum characteristics of noise neutron for PWR nuclear power plants of million kilowatt is analyzed with different power and different fuel cycles. The results show that it can recognize the internal vibration behavior of PWR nuclear power plants through the analysis of power density spectrum characteristics, which can provide a basis for internal state analysis of PWR nuclear plants.

Analysis and Diagnosis of Problem of Larger Amplitude with Core Barrel Beam Mode in a Nuclear Power Plant

Luo Ting, Liu Caixue, Hu Jianrong, Yang Taibo, Jian Jie, Feng Jintao, Ai Qiong

2018, 39(3): 184-187. doi: 10.13832/j.jnpe.2018.03.0184

Abstract(199) PDF(148)

Abstract:
Based on the analysis of the ex-core neutron noise signal in multiple fuel cycles in Fangjiashan and Ningde Nuclear Power Plants, the beam mode vibration frequency and amplitude characteristics of the core barrel is obtained. These characteristics are applied to other nuclear power plants, and the problem of larger amplitude with barrel beam mode in a nuclear power station is found. The trend of beam mode vibration frequency and amplitude, the drift of frequency, and the growth rate of amplitude are analyzed and diagnosed. No obvious support deterioration in the barrel is found, and the condition under which the plant can continue to operate is given.

2018 Vol. 39, No. 3