核动力工程

Radiation Safety System and Its Design Coordination for Nuclear-Powered Ship

Lin Xiaoling

2019, 40(5): 1-5.

Abstract(535) PDF(530)

Abstract:
Radiation Safety is of great importance for the combat effectiveness of nuclear-powered ships, while overemphasis on radiation safety could influence the performance of carriers negatively. As a result, the optimization of radiation protection is the crucial task in the design of nuclear-powered ships. To achieve this goal, the system of radiation safety for nuclear-powered ships is constructed. The function, main problems to be considered, the control index, the relations and cooperation from inside and outside is analyzed.

Experimental Study on Heat Transfer of Supercritical Water in 2×2 Rod Bundles

Li Yongliang, Huang Zhigang, Wen Yan, Zhu Haiyan, Zang Jinguang, Zeng Xiaokang, Yan Xiao, Huang Yanping, Xiao Zejun

2019, 40(5): 6-12. doi: 10.13832/j.jnpe.2019.05.0006

Abstract(270)

Abstract:
Heat transfer experiments of supercritical water in 2×2 rod bundles were conducted based on Chinese Supercritical Water Cooled Reactor(CSR1000) fuel assembly design. The experimental parameters were as follows: the system pressure 23~25 MPa, the mass flow flux 680~1400 kg/(m2?s), and the heat flux 174~968 kW/m2. The experimental results indicated that the heat transfer performance of the bundles reduced as heat flux increasing and mass flow flux decreasing while kept insensitive to the pressure as which varied from 23 MPa to 25 MPa. Moreover, it was found that the heat transfer performance of 2×2 rod bundles was influenced by both the differences from bulk flow to boundary layer in thermal physical properties and the non-uniform of flow and heat transfer in different sub-channels. Based on the experimental data, a satisfactory heat transfer correlation of 2×2 rod bundles was obtained, and about 88.9% of the experimental points had deviations from the correlation within ±25%.

Study on Thermal Hydraulic Characteristics in Startup of SCWR

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

2019, 40(5): 13-17.

Abstract(278)

Abstract:
To investigate the startup characteristics of whole supercritical pressure water cooled reactor (SCWR) system, the complete startup system model of the SCWR was established with the analysis code SCTRAN, based on the CSR1000 core, high performance light water reactor (HPLWR) steam cycle and SCWR circulation startup loop. The correctness of the model was verified in comparison with the steady-state parameters of the steam cycle of the HPLWR. A sliding pressure startup procedure with the circulation loop that employs a control system was analyzed, and the transient performances of the core, steam drum, steam turbine, reheaters, steam extraction, and heaters at each stage were obtained. The calculation results show that the startup sequence and startup thermal parameters agree well with the expectation: the system starts up stably and the core remains in the single phase; the inlet steam of the turbine stays supercritical; the core inlet temperature can reach 280℃ after the high-pressure and low-pressure heaters; the inlet pressure of the high-pressure turbine can be kept constant. During the startup procedure, the maximum cladding surface temperature remains below the limit temperature of 650℃. The entire startup procedure is safe and reliable.

Development and Verification of Thermal-Hydraulic Transient Analysis Code in Plate-Type Fuel Nuclear Reactor

Liu Wei, Zhang Yong, Jiang Xiaowei, Zhang Cheng, Zhang Dalin

2019, 40(5): 18-22.

Abstract(383)

Abstract:
The mathematic models for the thermal-hydraulic transient analysis of the plate-type fuel reactor system are established, and a code named SYSTRAN is developed. The design condition of China advanced research reactor (CARR) and the blockage transient case of the IAEA MTR are calculated to verify the SYSTRAN code. Comparison with the flow distribution, outlet temperature etc. obtained by SYSTRAN with the reference parameters indicates that the calculated results agree well with the reference values, which demonstrates that the SYSTRAN code can be used for the thermal-hydraulic analysis in the plate-type fuel reactor system.

Design Study of Thorium Molten Salt Reactor Body on Basis of Multi-Specialty Coupling Analysis

Yang Hongrun, Li Xiang, Yang Licai, Sun Yingxue, Zhang Zhuohua, Ying Dongchuan, Fu Qiang

2019, 40(5): 23-28.

Abstract(324)

Abstract:
Based on the characteristics of TMSR-SF1, the reactor body design and optimization method via multi-specialty coupling analysis was proposed in this paper. Referring to existing mature design codes and combining structure and function requirements of TMSR-SF1 reactor body, the structure drawing for the reactor body was presented firstly, then shielding design analysis, heat transfer and temperature field analysis, and mechanical analysis were carried out. Finally, the reactor body design of TMSR-SF1 has been preliminarily realized by using the multi-specialty coupling repeated iteration analysis, which satisfied the function of TMSR-SF1. Furthemore, reactor structure material selection demonstration and manufacturing feasibility analysis were also performed to ensure the engineering implementability of the structure design.

Numerical Analysis on the Characteristics of Steam Condensation in Presence of Air under Vertical Tube Bundle Conditions

Quan Biao, Bian Haozhi, Ding Ming, Luo Hanyu, Zou Zhiqiang, Li Feng, Sun Zhongning

2019, 40(5): 29-34.

Abstract(187)

Abstract:
Numerical simulations on steam condensation in the presence of air under tube bundle conditions were performed based on the software STAR-CCM+. Calculations were carried out based on a the 3 by 3 tube bundle which has a tube pitch of double diameter, aiming at recognizing local field distributions and thermal-hydraulic characteristics of various tubes. Results indicate that the air layers of various tubes interfere in the tube bundle region, forming a larger high concentration air layer region. On the one hand, this enlarges local velocity and facilitates convective heat transfer; and on the other hand increases the air layer thickness and inhibits condensation heat transfer. On the effect of bundle structure, the concentration, temperature and velocity in the tube bundle region are obviously different from those of the single tube one, sharply decreasing the local heat transfer coefficient by 50% and has a maximum variation of 1.88 times circumferentially. The axial heat transfer property mainly is affected by the development of concentration boundary layer, and circumferential heat transfer property by the adjacent tubes. By analyzing the average heat transfer coefficient, it indicates that compared to the single tube, the maximum reduction of the bundle tube average heat transfer coefficient is 9.06%.

Study on Coolant Flow Characteristics for Residual Air Criteria Improvement of Primary Loop Dynamic Venting

Zhang Zhao, Sheng Guolong, Sun Kaibao, Zhao Fuyu, Chong Daotong, Yan Junjie

2019, 40(5): 35-40.

Abstract(139)

Abstract:
According to the improvement proposal for the residual air criteria of the primary loop dynamic venting, the pressure prediction method for the main primary nodes and the judgment method for the two-phase coolant flow are investigated. The thermo-hydraulics model for the primary loop and its auxiliary systems are proposed. When the residual air criteria is raised to 24 normal cubic meters, the startup transient process of reactor coolant pumps are simulated with the models, to obtain the pressure of the main nodes in the primary loop. Based on the gas resolution and released model, the change trend of nitrogen saturation solubility and oxygen saturation solubility are obtained. The results indicate that the pressures of the main nodes are within the normal operation limit; the nitrogen and oxygen cannot release from the coolant and cannot form a two-phase flow. Therefore, in terms of the coolant flow characteristics, the improvement for the residual air criteria to 24 normal cubic meters is feasible.

Numerical Analysis on the Characteristics of Steam Condensation in Presence of Air under Vertical Tube Bundle Conditions

Jiang Hong, Zhou Jing, Liu Lizhi

2019, 40(5): 41-45.

Abstract(175)

Abstract:
In this paper, the impeller of the reactor coolant pump is analyzed by Computational Fluid Dynamics (CFD) method to obtain its hydraulic performances. According to the CFD analysis results, the impeller cannot match the system requirements and need be optimized because of the lower head under rated flow, the high flow incidence at the inlet of the blade around the region near the hub, and the high net positive suction head (NPSHr). Optimization targets are determined according to the calculation tolerances and engineering experiences. The impeller inlet angle and outlet angle, and the location of the head of blade are determined as variables of the optimization. After calculation and comparison, the final optimization proposal is determined. The analysis results for the optimized impeller show that the flow incidence at the inlet of the blade is decreased obviously, anti-cavitation capability and internal flow conditions in the impeller are better than the original one, and the optimization target is achieved.

Prediction of Critical Heat Flux Based on Gaussian Process Regression

Jiang Botao, Huang Xinbo

2019, 40(5): 46-50.

Abstract(173)

Abstract:
Accurate prediction of critical heat flux(CHF) is very important for reactor safety and operation. Aiming at the shortcomings of existing artificial neural networks(ANNs) prediction methods, a Gaussian process regression(GPR) based CHF prediction method is proposed. Firstly, the obtained CHF data under local conditions are preprocessed, and the data is divided into a training set and a test set. Then, the training data is used to train the GPR model, and the optimal hyper-parameters are obtained. Secondly, the trained GPR model is used to predict the CHF, and the results are compared with the radial basis function neural networks(RBFNN). Simultaneously, the effect of important parameters on CHF is analyzed. The results show that the prediction results of the GPR model have higher prediction accuracy and smaller error compared with RBFNN, and agree well with the experimental data. The parametric trends fit the general understandings.

Numerical Study of Fast Reactor Steam Generator Based on Porous Media Model

Wang Hongyang, Ruan Shenhui, Wen Qinglong, Chen Zhiqiang

2019, 40(5): 51-55.

Abstract(257)

Abstract:
As a heat exchanger between the primary side sodium and secondary side water, the steam generator (SG) is a crucial equipment ensuring the safe operation of the reactor. Thus it is essential to analyze the hydraulic characteristics in primary side of SG. In this paper, a numerical investigation of three-dimensional thermo-hydraulic characteristics in the primary side of a fast reactor steam generator is simulated based on the porous media model. The momentum source term for the porous media model is obtained by investigating the flow characteristics in tube support plates (TSPs) and tube bundle region. The heat transfer from secondary coolant to primary sodium is calculated by RELAP5 code, and the energy source term is compiled into FLUENT solver by user defined functions (UDF). The thermo-hydraulic characteristics of the SG are obtained by solving the governing equations by ANSYS FLUENT solver. The accuracy of the simulation is verified by comparing with the designed data.

Optimization of Core of Advanced High-Temperature Reactor Based on Hybrid Adaptive Genetic Annealing Algorithm

He Liaoyuan, Xu, Yan Rui, Zou Yang, Guo Wei, Liu Guimin

2019, 40(5): 56-60.

Abstract(357)

Abstract:
The predesigned Plate-type Advanced High Temperature Reactor (AHTR) used the uniform enrichment fuel assemblies, which resulted in the high power peak factor with its total power peak factor of 2.09. Thus, the safety and economy of the reactor are restricted to a certain extent. In this paper, different enrichment fuel assemblies were used to flatten the power. In order to speed up the search of the optimal loading pattern of the reactor core, an adaptive intelligent algorithm was adopted. The whole optimization process were completed by an automatically program written by MATLAB. The optimized power peak factor is reduced to 1.122, which is reduced by 25.02% compared with the original design. Simulation results of the temperature field show that the temperature distribution of the optimized reactor is more uniform. The maximum temperature is decreased from 1030 K to 1010 K, which greatly increases the temperature margin of the core.

Numerical Analysis of Bundle Effect on Steam Condensation in Presence of Air

Quan Biao, Bian Haozhi, Ding Ming, Li Yi, Cheng Xiang, Peng Hang, Sun Zhongning

2019, 40(5): 61-66.

Abstract(199)

Abstract:
This paper numerically investigated the tube bundle effect on steam condensation in the presence of air. Simulations were firstly performed based on the 3 by 3 tube bundles with various tube pitches. Via these cases, the tube bundle effect was defined and the effects of tube pitch on local and average condensation heat transfer were discussed. Then at a tube pitch of 1.5d, the effect of bundle structure on the condensation heat transfer was evaluated. Results indicate that the bundle effect includes the high concentration air layer inhibition effect and the bundle suction enhancement effect. With the decreasing of the tube pitch, the second and third type tubes are mainly influenced by the high concentration air layer effect, and the first type tubes mainly by bundle suction effect. At 1.5d tube pitch, the average heat transfer coefficient for the second and third type tubes decreased 6% and 29% compared to the single tube. In comparison, the first type tubes increased 2.5%. At the tube pitch of 1.5d, the bundle suction effect increases with the increasing of the tube column, resulting in an enhancement of bundle average condensation heat transfer coefficient. For the 3 by 20 bundle structure, the bundle average heat transfer coefficient exceeded that of the single tube.

Stability Analysis of Final-Stage Flow Field in Low Volume Flow Working Conditions for Marine Nuclear Wet Steam Turbine

Lu Yingdong, Yang Zichun, Zhang Lei, Cao Yueyun

2019, 40(5): 67-73.

Abstract(169)

Abstract:
Bladegen software is used to parameterize the solid twisted blade, and Turbogrid software is used to complete the accurate modeling and meshing of the complex twisted blade single. The final stage of the nuclear wet steam turbine in variable volume flow condition were calculated and analyzed by CFX software. The results show that the flow stability in the final stage of the wet steam turbine gradually decreases as the relative volume flow decreases. Steam flowing in the stage exhibits extremely complex rules, and in the 70%~85% leaf of height area, the flow field is the biggest and the stability is the worst. The research conclusion provides a reference for the future optimization of the structure and the blade shape of the steam turbine, and further improves the stage flow capacity.

Optimization of Power Ramp Rate in CPR1000 Reactor Restarting

Li Changzheng, Lin Shaofang, Liu Weichao, Cai Zhiyi, Nie Lihong, Deng Yongjun

2019, 40(5): 74-78.

Abstract(306)

Abstract:
Restart strategy applied in CPR1000 units can be further improved according to the operation experience and the simulation conducted by software. This paper simulates the restarting and full power operating modes of the reactor in terms of ramp rate and Tthreshold power level. Conditioning occurs after tens of hours of operation at full power. The capacity factor can be increased about 0.1% by optimizing the ramp rate.

Research of Main Influencing Factors during Sol-Gel Preparing Process of U-Zr-Hf Burable Poison Fuel

Zhang Jia, Li Jia, Kang Wu, Liu Yu, Peng Xiaoming, Liu Jinhong, Zeng Cheng

2019, 40(5): 79-84.

Abstract(210)

Abstract:
For the development of the integrated type burnable poison fuel with good irradiation stability, the U-Zr-Hf burnable poison fuel was prepared by sol-gel method, which can achieve the uniform distribution of burnable poison and fuel. Effect of the main influencing factors in preparation of broth, broth droplet gel-forming, gel particles washing and reduction sintering on preparation process and performance of fuel was studied. The results showed that: the total concentration of metal ions was the main reason for the viscosity increasing at low temperature. And the main factor for steady broth droplet gel-forming temperature was the metal ions system. The volume ratio of washing liquid and gel particles affected the washing process. The increasing of sintering temperature promoted the densifying of fuels. The metallographic surface photos of U-Zr-Hf fuel showed that the fuel was apparent integrity, with no rupture, and internal compactness.

Releasing Model of Fission Gas Based on Dispersion Fuel Particle Cracking

Chen Hongsheng, Long Chongsheng, Xiao Hongxing, Wei Tianguo, Gao Wen

2019, 40(5): 85-91.

Abstract(232)

Abstract:
According to three releasing approaches of fission gas after the cracking of the dispersion fuel particles, the releasing models of crack connection, bubble connection and atom diffusion were established respectively, resulting in the releasing model of fission gas based on the dispersion fuel particle cracking. The releasing amount of fission gas was calculated by this model. Results show that the releasing amount of fission gas is mainly attributed to the releasing approach of crack connection, and the increasing rate of releasing amount gradually increases with the burn-up. With the increase of annealing temperature, the releasing amount rapidly raises, while the increasing rate gradually decreases with the annealing time. The crack widths calculated from the releasing model of fission gas are consistent with the experimental data, which verify the releasing model of fission gas based on the dispersion fuel particle cracking.

Code Validation and Study on High Burnup Fission Gas Release Behaviors

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

2019, 40(5): 92-96.

Abstract(307)

Abstract:
Calculations and analyses of Halden IFA519.9 DK fuel rod irradiation test were carried out using the fuel performance code COPERNIC. Investigations on the high burnup fission gas release behaviors were implemented. Comparison between predictions and experimental results were performed. The results indicate that, during high burnup to ~100 GW?d/t(U) irradiation, the code predictions show good agreement with experimental data in terms of fission gas release. The code does not particularly model the evolution of fuel porosity within high burnup rim structure, however, the accuracy and rationality on fuel rod integrated performance analysis would not be affected.

Research of Modeling Methodology for Supporting System Initiating Events

Yang Jian, Wang Yuqing, Feng Churan

2019, 40(5): 97-102.

Abstract(207)

Abstract:
Support system initiating event (SSIE) is a special type of initiating event and should be treated appropriately in the probability safety assessment (PSA). The important technique issues to be addressed in modeling include: PSA modeling linking and quantification methodology, mission time for the standby trains or components, common cause failure (CCF), importance and uncertainty analysis results. The treatments method of these issues was revealed inconsistent for different analyzers and influenced the risk insights of PSA significantly at the present time. Based on the requirements of technical standards, this paper put forward the following suggestions through case analysis and comparison: a. SSIE fault trees should be linked with PSA overall model and quantified; b. Currently two common methods-Multiplier and Explicit are both available. However, the limitations of the two methods in the analysis of importance and uncertainty should be understood, and obvious deviations should be avoided.

Development and Application of Main Control Station of Nuclear I&C System Platform Firmsys

Shi Guilian, Zhou Fei

2019, 40(5): 103-107.

Abstract(191)

Abstract:
The development and the application of China’s first nuclear safety level digital I&C system FirmSys platform in the main control station (MCS） are conducted. PowerPC series high-performance processor is used on the hardware of the main controller unit. Non-commercial operating system, task scheduling without interruption and static fixed memory distribution design are used on the embedded software. Communication between two control stations uses the one-way transmission mode. MCS meets the design requirements of fail-safe in the event of a failure and loss of electric power. MCS software has passed the verification and validation (V&V) of the independent third party from German ISTec Organization. All the cards in MCS have passed the environmental, electromagnetic compatibility and seismic tests required by relevant standards, and have been successfully applied in ACPR1000 Nuclear Power Plant and HTR Nuclear Power Plants.

Design and Analysis of Fan Free Cabinet Heat Dissipation for Nuclear Safety Cabinet

Li Zhaolong, Zhang Yunbo, Jiang Zhirui, Liu Yongliang, Shi Yingbin

2019, 40(5): 108-110.

Abstract(209)

Abstract:
Currently, in the nuclear safety class DCS cabinet, forced air cooling is the main way to dissipate the heat. However, the life of fans is often shorter, which requires the real-time monitoring of their operation status and regular maintenance and replacement, thus increasing the use cost of the nuclear safety DCS cabinets. Compared with the forced cooling through fans, natural cooling without fans is a more reliable and economical method. This paper introduces a design and analysis method for the natural convection heat dissipation system, and verifies the rationality of the analysis method by Flotherm electronic heat dissipation analysis software. According to this analysis method, a cabinet without fan for nuclear safety DCS FirmSys is designed. The cabinet has been applied in the second phase reconstruction project of Qinshan Nuclear Power Plant.

Reason Analysis and Improvement Measures Evaluation for Water Intake Blockage at Northern Nuclear Power Plants

Zhang Guohui, Song Hehang, Mu Yangyang

2019, 40(5): 111-117.

Abstract(342)

Abstract:
In recent years, the water intake of a nuclear power plant in the north of Liaodong Bay, Bohai Sea, has been blocked several times by marine organism, resulting in the failure of the function of the cold source systems, and threatening unit safety. According to the actual situation of each water intake blockage event in the Nuclear Power Plant, the reason of blockage is analyzed, meantime the improvement measures and efforts in cold source guarantee system establishment and the control strategy of the Nuclear Power Plant are investigated and discussed. Finally, t the pending problems of the cold source is summarized, and the optimization in terms of water intake structure, how to cope with bad weather, strengthening monitoring and early warning are proposed,, to enhance the response capability of the Nuclear Power Plant to the outbreaks of marine organisms, and reduce the risk of water intake blockage.

Effect of Inclination Condition on LOCA for a Small Offshore Reactor

Cao Zhiwei, Liu Jianchang, Xiao Hong, Yang Jiang, Lu Xianghui, Tian Wenxi

2019, 40(5): 118-123.

Abstract(164)

Abstract:
Based on the thermal-hydraulic system analysis code RELAP/SCDAPSIM, the models of the primary loop and secondary loop and the safety injection system of the small offshore reactor are built, and LOCA with a double ended break under different inclination angles are simulated. Simulation results show that, longitudinal inclination has little effect on thermal parameters of the reactor, while lateral inclination does have a significant influence, and steep edge effect exists. Particularly, when the lateral inclining with large angle occurs, there is a significant reduction of core water level due to the redistribution of liquid coolant in the primary loop, and peak cladding temperature of the fuel is raised about 520℃.

Experimental Study on Chemical Effects of Nuclear Power Plant Sump Strainer

Liu Weiwei, Xia Xiaojiao, Ma Weigang, Jiang E, Fu Shengwei, Zhao Yongfu, He Yanchun

2019, 40(5): 124-129.

Abstract(173)

Abstract:
Under the loss of coolant accident in nuclear power plants, the blockage of sump strainers may be resulted from chemical effects, which will affect the recirculation function of the emergency core cooling system or the containment spray system, disabling the cooling of the core and containment, and threatening the safety of nuclear power plants. Study on the effects of potential chemical products on sump strainer head loss after the loss of coolant accident in Qinshan PhaseⅡExtension Project was conducted. The results showed that the alumina-bearing material and the insulation material released aluminum and silicon in the sump environment, and that aluminum and silicon formed chemical precipitation and deposited on the debris bed. This can result in the blockage of the debris bed and the reduction of the porosity, and increase the head loss of the sump strainer. There are chemical effects after loss of coolant accident in Qinshan Phase Ⅱ Extension Project, and it should be considered in the sump strainer performance evaluation and downstream effect analysis.

Simulation of Radiation Condition in HWR NPPs under a Postulated Severe Accident

Wang Xuedong, Cao Xuewu, Zhao Xiaoling

2019, 40(5): 130-134.

Abstract(210)

Abstract:
the MCNP code is used to simulate the radiation environmental conditions of different regions of the power plant at different periods under a postulated accident. The results indicate that peak dose level appeared about 2.5 h after the release of the nuclides in the NPPs areas, and thereafter, the dose level gradually decreases as the decay of the nuclide proceeds; in addition, the areas near the ECCS equipment and pipelines maintain an extremely high dose rate for a long time after accidents, and the expected personnel accessibility problem will produce a threat to accident mitigation.

Numerical Study on Effect of Ocean Conditions on Jet Pump Cavitation Characteristics for Marine Nuclear Power Plants

Li Yong, Wang Wei, Yao Shiwei, Wei Wei, Xiao Qi, Li Shaodan

2019, 40(5): 135-139.

Abstract(163)

Abstract:
Owing to the limited space in the marine nuclear power plants, the lifting height for the condensate system is quite short, which can induce the cavitation of the condensate pump. To solve this problem, a jet pump is mounted in front of the condensate pump. Nevertheless, the cavitation risk also exists on the jet pump in the harsh ocean environment such as rolling motion. Therefore, the cavitation characteristics of the jet pump are investigated by CFD numerical method under the stable condition as well as two rolling motion condition. The results show that under the stable condition, the void fraction inside the jet pump flow field is very small, and the nozzle exit and throat pipe forepart are two positions prone to cavitation. Ocean rolling motion results in the enhancement of the cavitation of the jet pump and the decreasing of the lift range. The more frequent the rolling, the greater the effect. Especially, when the rolling motion period is 3 s, the cavitation of the jet pump develops seriously, the pump lift surges and is less than the rated lift evidently, which affect the operation safety of the condensate system.

Cross-Scale Refined Damage Evolution Analysis of Impact of Large Commercial Aircraft on a Reactor Building Based on Octree-SBFEM

Zou Degao, Sui Yi, Chen Kai, Pan Rong, Xiong Jingchuan

2019, 40(5): 140-145.

Abstract(171)

Abstract:
Combined with the scaled boundary polyhedron finite element analysis method, this paper firstly applies the Octree Technique to the nuclear engineering. The refined damage evolution analysis of large commercial aircraft crashing into Generation Ⅲ+ nuclear power plants is developed. Meanwhile, the influence of foundation effect, shape of impact region chosen and SSI effect is discussed. The results indicate that the cross-scale refined analysis method is with extremely strong capability of discrete grid, high quality and little quantity of element. Furthermore, the method is highly flexible for the model modification; the refined FEM model simulates more accurately the damage evolution and gradually destruction process; the structure-soil interaction (SSI) effect in the analysis of the aircraft impact on nuclear islands cannot be ignored in non-lithology foundation.

Development of Underwater Decontamination Device for Spent Fuel Storage Rack in Nuclear Power Plants

Luo Wenguang, Ou Jianle

2019, 40(5): 146-149.

Abstract(116)

Abstract:
Aiming at the necessity and functional requirements of the decontamination of the spent fuel storage rack, a underwater flushing decontamination device was developed for the spent fuel storage rack in a second-generation nuclear power plant. The structural composition, functional principle and control system design of the device are introduced in detail. The device is easy to operate and has good flushing and decontamination capability, which greatly reduces the radiation dose level of the spent fuel storage rack.

Development of Inspection System for Weld Joint between Steam Generator and Reactor Coolant Pump in Nuclear Power Plants

Yu Gang, Zhou Lusheng, Tao Zeyong, Zhang Baojun

2019, 40(5): 150-155.

Abstract(147)

Abstract:
The reactor coolant pump casing of weld joint between steam generator outlet nozzle and reactor coolant pump is made by coarse-grain austenitic material. The weld joint is with thick wall, ultrasonic attenuation and grain scattering, the development of ultrasonic technique is very difficult. In this study, a special design is adopted to develop an automatic ultrasonic inspection system from inner surface of the steam generator outlet nozzle, and this system has been applied to the pre-service inspection of AP1000 Nuclear Power Plant in China. The results show that this system could fully meets the on-site inspection requirements, and the inspection results are in good agreement with the factory results.

Study on Measurement of Steam Generator Water Level in Floating Nuclear Power Plants

Zhao Yang, Lyu Xin, Zhu Biwei, Zheng Songhua, Wu Qian, Wang Xuemei, Wan Yi, Luo Hanyu

2019, 40(5): 156-159.

Abstract(100)

Abstract:
The differential level measurement technology used in land nuclear power plants (NPPs) cannot be directly applied in the measurement of steam generator (SG) water level of floating NPPs because of the long-term inclination and swing condition. In this paper, combined with the principle of liquid level measurement by differential pressure, SG water level measurement improvement scheme for floating NPPs is proposed based on theoretical calculation and analysis. This improvement scheme was tested and verified by inclination and swing test bench. The related experimentation verifies that this improvement scheme can track the water level under various working conditions and get the accurate water level under inclination and swing condition.

An Offline Debugging Tool of Nuclear Security Protection Algorithm in the DCS

Zhang Chunlei, Zhang Baoqian, Ren Baohua, Peng Li, Zhang Zhihui

2019, 40(5): 160-164.

Abstract(192)

Abstract:
During the engineering design of DCS for the nuclear power plants, it is difficult to find configuration errors such as protection algorithm logic and human errors by manual check, and it is also hard for the designers to evaluate and analyze the dynamic characteristics. It is inefficient to debug the algorithm through connecting devices, and the devices cannot support functions such as suspension, back, fallback and jump, which leads to lacking of efficient methods to locate errors. It is an effective means to use offline debugging tools to debug protection algorithm in the safety DCS of nuclear power plants. Based on FirmSys, the first nuclear safety level DCS, which China owns the intellectual property rights, an offline debugging tool in nuclear safety protection algorithm has been designed in this paper, which realizes the debugging of logic control protection and can meet the requirements of nuclear power standards. This method can improve the correctness and rationality of the logic configuration. The tool can shorten the design cycle and increase the debugging efficiency of nuclear safety protection algorithm, and it also provides a fast and efficient way for field debugging and locating errors. It has been used in several nuclear power projects, such as Yangjiang Nuclear Plant Plant and Hongyanhe Nuclear Power Plant.

Analysis of Operator's Task under Rapid Power Change in Nuclear Power Plants

Liu Xueyang, Zhang Li, Zou Yanhua

2019, 40(5): 165-169.

Abstract(122)

Abstract:
The differences between the peak load and the operation task in the normal operating conditions of the nuclear power plants are compared from three aspects of operator training, task type and working load. Compared with the operation tasks under the normal working conditions, the mental load and the physical load of the operators would be changed greatly, which would lead to the difference between the operator's cognitive model and the human error mode compared with the conventional working condition. The existing HRA methods and models are difficult to satisfy the operator's reliability analysis, so a new HRA method is needed to solve the problem of the reliability under the background of continuous and fast changing of control tasks in the nuclear power plants.

Application of FMEA Technology in CCM Equipment Maintenance Optimization of CPR1000 Unit

Yang Lifei, Wu Tao, Qing Chen, Liu Xiaolei, Shi Lei

2019, 40(5): 175-179.

Abstract(476)

Abstract:
In this paper, the failure mode and effect analysis (FMEA) method and process are introduced. Combined with the maintenance management characteristics of nuclear power plants, a simplified FMEA method is formed. The failure mode, fault impact and maintenance strategy of the critical component management (CCM) equipment in CPR1000 units are analyzed. The CCM equipment FMEA database is established. Practice has proved that the development of this work not only identifies the potential shutdown and shutdown failure modes that CCM equipment management has not covered, but also finds errors and inconsistencies in the CCM equipment technical documents, and also reviews the correctness and completeness of the CCM equipment list. At the same time, based on group operation and maintenance big data, the maintenance strategies of multiple CCM equipments are systematically optimized to make up for the lack of maintenance management of CCM equipment, reduce unnecessary investment in CCM equipment maintenance resources and manage equipment related to CCM equipment. It is of important reference value for the implementation of the management of CCM equipment and the mitigation of the risk of unplanned shutdown of nuclear power plants.

Effect of Back Pressure of Containment on SGTR Accident Process

Jiang Xiaowei, Deng Jian, Qiu Zhifang, Zhu Dahuan, Dang Gaojian, Zhang Dan, Bi Shumao

2019, 40(5): 180-183.

Abstract(184)

Abstract:
In the advanced PWR SGTR accident, the passive residual heat removal system was designed to remove the decay heat in the RCS with a heat exchanger which is immersed in the RWST. The back pressure of the containment used in the analysis will affect the temperature of the RWST water boiling and the temperature difference between the two sides of the heat exchanger, thus affect the heat transfer efficiency. In this paper, the increasing of the containment pressure and temperature caused by the RWST water boiling is analyzed to determine the containment pressure process during the SGTR accident. A comparative research on the accident process under different containment pressure was conducted to determine the effect of the containment pressure on the accident process. The analysis shows that the higher the back pressure of the containment, the smaller the temperature difference between the two sides of the heat exchanger, and the weaker the heat transfer capacity of the passive residual heat removal system. The higher back pressure will prolong the accident process and the end time of the broken flow, increase the amount of coolant released, and decrease the margin of the overflow.

Research and Design on First of a Kind Test of HPR1000

Huang Zongren, Liu Changwen, Lai Jianyong, Li Feng, Wang Xiaoyu, Li Yan, Li Haiying, Leng Guijun

2019, 40(5): 184-186.

Abstract(228)

Abstract:
This paper introduced the NNSA, IAEA and NRC requirements for first of a kind test of nuclear power plants, and determined the design philosophy of the first of a kind commission test of HPR1000 combined with the operation experience feedback of nuclear power plants and engineering practice of similar nuclear power plants. The first of a kind test items of HPR1000 is researched and determined, based on the analysis of the new design concept and new design feature used in HPR1000. The test condition, test content and test acceptance criteria of the first of a kind test items were analyzed one by one, in order to guide the First of a Kind Test of HPR1000.

Benchmark Validation of Resonance Calculation in Advanced Neutronics Lattice Code KYLIN-Ⅱ

Tu Xiaolan, Chai Xiaoming, Lu Wei, Chen Dingyong, Guo Fengchen, Yin Qiang, Tang Chenhang

2019, 40(5): 187-191.

Abstract(191)

Abstract:
Several benchmark problems were validated for the resonance calculation module of neurotics lattice code KYLIN-Ⅱ, including cell problem, IAEA problem, thorium based assembly problem, multi-layer bushing fuel assembly problem, burnable poison lattice problem, supercritical water reactor fuel lattice with large water cavity, and AFA3G fuel assembly with Gadolinium problem. The validation results show that the resonance calculation module in this paper is suitable for complex geometric structures such as square assembly with fuel rod lattice and planar fuel assembly and hexagonal assembly. Besides, it can correctly calculate the resonance problem of complex materials containing uranium, thorium, poisons, etc. It satisfies the requirement of future engineering application.

Experimental Study on Flow Induced Vibration of Different Size Strips of Fuel Rod Bundles

Zhang Botao, Zhu Yechen, Gong Shengjie, Gu Hanyang

2019, 40(5): 192-196.

Abstract(160)

Abstract:
To thoroughly study the vibration characteristics of the grid strip of fuel rod bundles, experimental study on flow induced vibration of different size flat strips was conducted, and the vibration characteristics of the strips were acquired. Within the range of experimental tests, the vibration response of the strips could be divided into two parts: turbulence induced vibration and vortex induced vibration. The predominant vibration response induced by turbulence was the first mode of the strips, and the response was amplified with the increasing of the flow velocity. For vortex induced vibration, the St number was from 0.2 to 0.25 in this study, and the lock-in phenomenon was observed. For a strip, the lock-in range of the lower modes was larger than the higher modes. For different thickness strips, the lock-in range of the same modes narrowed down with the increasing of the thickness. For different length strips, the lock-in range of the long strips was larger than that of the short strips of the same modes.

Numerical Investigations of Supercritical Hydrogen Flow and Heat Transfer in Curved Tube

Zhou Biao, Ji Yu, Sun Jun, Sun Yuliang, Shi Lei

2019, 40(5): 197-201.

Abstract(186)

Abstract:
In order to investigate the flow and heat transfer characteristics of the supercritical hydrogen running through the nozzle throat, this paper simulated the supercritical hydrogen flow in a 180° curved tube by ANSYS FLUENT, and obtained the velocity profiles and the wall temperature at various interested positions. It was found that the centrifugal force forced the hydrogen flow shift radially outward in the bend and induced the secondary flow perpendicular to the main flow, making the flow velocity on the inner side lower than that of the outer side. Furthermore, because of the non-uniform flow distribution along the circumferential tube wall, the heat transfer on the outer wall of the bend was strengthened, while deteriorated on the inner side. Based on the operating conditions in the current paper, the inner wall close to the exit of the curved section has the highest wall temperature, with the most notable heat transfer deterioration.

Calculation of Stress Intensity Factor of Zr-4 Cladding Tube by Flexibility Method

Chen Liang, Song Xiaoming, Pang Hua, Wang Kecheng, Huo Meng, Zeng Xiangguo

2019, 40(5): 202-206.

Abstract(219)

Abstract:
Based on the virtual crack closure technology (VCCT), by using the method of single variable control changes, the numerical model is established in the light of the symmetric double edge crack in thin-walled cylindrical shell by symmetrical concentrated force. The crack tip energy release rate is obtained. The effect of the Zr-4 alloy cladding tube prefabricated crack length and external load on the stress intensity factor is analyzed. The approximate formula of stress intensity factor (SIF) was obtained by using the flexibility method. The validation results are in good agreement with the numerical results.

2019 Vol. 40, No. 5