核动力工程

The Research on Burnup Characteristic of Doping Burnable Posion in PWR

QIANG Shenglong, QIN Dong, CHAI Xiaoming, YAO Dong

2014, 35(2): 1-4.

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In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons（such as natural element, manual nuclide and soluble boron） in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf、Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core.

Development of Nuclear Reactor Core’s Full Offset Refueling Algorithm Based on Two-Dimensional Array

ZHAO Apeng, LI Yuanhao, WU Fengqi, LU Xiusheng

2014, 35(2): 5-7.

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The reactor’s refueling algorithm, which determines safety and efficiency of whole core’s refueling progress, is the key part of manipulator crane control system. This paper presents a new kind of full offset refueling algorithm through reactor core’s modeling based on two-dimensional array. It also resolved significant issues related to the algorithm. The engineering practice shows that the new algorithm can save refueling time prior to ensure nuclear safety, which means bringing considerable benefits to current operating nuclear power reactor station.

Characteristics Study for Decay Heat and Fission Product Inventory of Reactor Core with MOX Fuel

TAN Yi, WEI Shuping, DENG Lilin, LIU Xiaoli

2014, 35(2): 8-12.

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The MOX fuel has a good future for widely application. Today, the MOX fuel technology has been developed and applied in many PWRs all over the world. But there is still no experience in China and a lot of analysis works are needed. In this paper, the decay heat of the reactor core, heat load of spent fuel pool and fission product inventories of the reactor core according to the fuel management of M310 NPP with one third of the core filled with MOX fuel assemblies are studied. Comparing the reactor core with MOX assemblies with the reactor core with UO₂ assemblies, it shows that the changes of the decay heat of the reactor core, heat load of spent fuel pool are small, and they would not be 15% greater than UO₂ fuel. The change of fission product inventories is also not large except some fission product nuclides as 135Xe and 136Cs, which are over 40% more than UO₂ fuel. The result of this study proves that the radiation characteristics of MOX fuel core changes small, and has negligible impact on the plant safety and operation.

Calculation and Experimental Validation of ⁹⁹Tc、¹²⁹I Transmutation Rate in Xi’an Pulsed Reactor

WANG Lipeng, JIANG Xinbiao, ZHAO Zhumin, CHEN Lixin, LI Xuesong, XU Jiang, WU Hongchun

2014, 35(2): 13-16.

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The theoretical simulation and experimental methods related to Long Lived Fission Product（LLFP） of ⁹⁹Tc and ¹²⁹I transmutation in nuclear spent fuel were studied in order to develop the feasibility research on ⁹⁹Tc and ¹²⁹I transmutation in XAPR reactor. ⁹⁹Tc and ¹²⁹I ACE format cross libraries in XAPR were generated by NJOY code based on ENDF/B VII.0 library, different parameters were analyzed. MCNP code was used to modify ⁹⁹Tc and ¹²⁹I capture cross sections by adopting new generating ACE format and CINDER’90 cross sections of 63 groups. Transmutation rates’ calculations of ⁹⁹Tc and ¹²⁹I in XAPR were done by ORIGEN2 code, as well as comparison with experimental data, the difference between theoretical and experimental data mainly comes from the deviation of neutron flux used and the actual one. Conclusions were made that XAPR is effective in transmutation of ⁹⁹Tc and ¹²⁹I.

Study on Analysis Methods and Characteristics of Large Break Loss of Coolant Accident of Reactor Cores with Tight Lattice Design

HUANG Daishun, FU Ran, SHEN Yaou, WU Dan, YU Hongxing

2014, 35(2): 17-20.

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A key problem related with the tight lattice core design is whether it could be cooled successfully in a loss of coolant accident（LOCA）. In this paper, the modified code RELAP5/TIGHT was used to calculate and analyze the characteristics of large break loss of coolant accident of reactor cores with two different tight lattice design. It was found that compared with standard reactor core designs, the transient progress of a LOCA accident lasted a longer time for a tight lattice core design; the peak cladding temperature in the reflooding phase was much higher than that in the blown-down phase; the ratio of center distance between fuel rods to outer diameter of fuel rod（p/d） has a great impact on the peak cladding temperature; the increasing in the power density results in the increasing need to the safety injection flowrate. For safety consideration, p/d ratio less than 1.10 is not recommended.

Improvement of Sodium Boiling Model in Accident Simulation

ZHAO Shufeng, JIANG Hua

2014, 35(2): 21-26.

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To carry out the accident analysis of Chinese Experimental Fast Reactor, total instantaneous blockage of single subassembly（in short as TIB） is selected as analyzing object. Meanwhile BE+1 experiment of SCARABEE series experiments is selected as validation experiment. During TIB simulation, two-fluid six-equation model is used in sodium boiling period, and the subassembly channel is discretized by intercrossing discretion method radially and axially. In the course of model solution, a optimizing method is put forward aiming at solving some drawback of sub-channel model, which is on the basis of Lame computation factor, and the result is validated on BE+1 experiment, which show a reasonable outcome.

Sub-Channel Analysis of SCWR Assembly with Double-row Fuel Rods

CHEN Chao, DAN Jianqiang, ZHANG Bo

2014, 35(2): 27-32.

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In the present paper, an advanced sub-channel analysis program, ATHAS, was used to analyze double-row fuel rods assembly of a Supercritical Water Cooled Reactor（CGN-SCWR）, to inspect whether the thermal parameters such as fuel rod cladding temperature meet the safety requirements. Some representative groups of sub-channel parameters depending on both hydraulic diameter and outlet temperature were employed to more detailed study the impact of components on different types of sub-channel. Furthermore, in order to better study the effect of empirical correlations on the heat transfer of cladding, turbulent mixing coefficient, axial friction coefficient and heat transfer correlations were also chosen to make the sensitivity analysis. Results show that all the designed parameters satisfy the safety requirements, and the maximum cladding temperature reached to 685.3℃. The conclusions also suggest that the heat transfer effect was mostly effected by the chosen empirical correlations, for example the maximum cladding temperature difference for different correlations could reach to 41.3℃.

Numerical Simulation of Heat Transfer Characteristics of 2×2 Wire-Wrap Rod Bundles under Supercritical Conditions

ZANG Jinguang, YAN Xiao, HUANG Shanfang, HUANG Yanping, YU Junchong

2014, 35(2): 33-36.

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The SCWR assembly always adopts the tight lattice configure with the wire wrap to fix up. The wire wrap will contribute the turbulent mixing in different subchannels and have big impact on the heat transfer characteristics in the rod bundles. This paper took the small wire-wrap 2×2 rod bundles as the analysis object and investigated the heat transfer phenomena with the CFD tools. It was found that the wire wrap will enhance the heat transfer capability and impair the non uniformity of wall temperature. However, it could also bring about the local wall peak temperature which could be classified into two reasons. The impact of wire wrap structure parameter on heat transfer was also explained.

Three-Dimensional Steady Calculation on Two-Phase Flow in Secondary Side of Steam Generator

CONG Tenglong, TIAN Wenxi, QIU Suizheng, SU Guanghui, XIE Yongcheng, YAO Yangui

2014, 35(2): 37-40.

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Secondary side two-phase flow in steam generator was simulated based on the porous media model. Additional momentum and energy source terms were appended to the momentum and energy equations of porous media region, respectively. The additional momentum source contained the resistances of downcomer, tube bundle, support plate and separator. The additional energy source included the heat transfer from primary side to secondary side fluid. Solving the control equations by ANSYS FLUENT solvers yielded the distributions of velocity, temperature, pressure, density and quality. The distributions of velocity, void fraction and mixture density in hot side vary significantly from these in cold side; the void fraction at the inlet of separator differs from 0.63 to 0.98; pressure descends along axial direction, while at the positions of support plates, pressure declines sharply.

Research on Model of Flow Instability in Parallel Rectangular Channels

QIAN Libo, DING Shuhua, QIU Suizheng

2014, 35(2): 41-46.

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A mathematical and physical model for flow instability in one dimensional parallel rectangular channels was established, including the entrance section, heater section and riser section, and the code for flow instability in parallel two-rectangular channel was developed based on the mathematical model, which was validated with the experimental results of flow instability. Based on the code, the effect of the system pressure, the inlet and outlet throttling on the margin of stability and the frequency of oscillations was analyzed. The results showed that the margin of stability and the distribution of frequency were overlapped at different system pressure and the system would be stabilized and the frequency of the system would increase with the increasing of the inlet throttling and system pressure and the decreasing of outlet throttling.

Bubble Dynamics Behavior of Nanofluids to Enhance the Capability of In-Vessel Retention

WANG Yun, WU Junmei

2014, 35(2): 47-52.

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In this study, regression analysis on Al₂O₃/H₂Onanofluid properties was made. The growth and departure of the bubble behavior in the Al₂O₃/H₂Onanofluid and pure water flow boiling process were numerically simulated based on Moving Particle Semi-implicit（MPS） method in different conditions. And the comparative analysis was made between the results of pure water and Al₂O₃/H₂Onanofluid under the same conditions. The results indicate that the bubble in Al₂O₃/H₂Onanofluid grows faster and the bubble departure frequency of nanofluids is greater than that of pure water, and the heat flux is also improved in the same conditions. It is convinced that nanofluids can enhance the boiling heat transfer. This work also initially reveals the underlying mechanism why nanofluids can enhance the boiling heat transfer from the point of view of bubble dynamic behavior. In order to investigate the nanofluids application in In-Vessel Retention（IVR） strategy after severe accident happened to the nuclear reactor, the bubble dynamic behavior has been researched. The results indicate that the bubbles in nanofluids can depart from the heat surface earlier than that of the pure water, and this makes the new cooling fluid fill the heat surface timely and enhance the IVR capability.

Research of Phase Separation Phenomenon in Two-phase Bubbly Flow in T-Junction

WANG Laishun, LIU Lifang, TIAN Wenxi, MENG Zhaoming, YANG Yanhua, SU Guanghui, WANG Chenglong

2014, 35(2): 53-57.

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In this paper, we use CFD software ANSYS CFX to simulate the separation phenomena of two-phase（air-water） flowing through branch pipe when the flow pattern in the horizontal pipe is mainly bubble flow for T-junction of AP1000 consisting of ADS-4 and primary loop. The separation ratio, phase profile, pressure profile and velocity profile are obtained, and the effect of different inlet volume fraction and bubble size on the characteristics of separation is also obtained. The results indicate that the phase separation phenomenon at the T-junction position is significant, separation ratio will decrease with an increase of the inlet volume fraction, and there exists a bubble size that makes the phase separation effect most significant for a particular pipe.

Wall Effect Analysis Based on Subchannel Decomposition Method

ZANG Jinguang, YAN Xiao, HUANG Shanfang, HUANG Yanping

2014, 35(2): 58-62.

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When doing experiments with repeated structure elements, it is a routine to select part of it due to the limit of experimental cost and condition. This part of structure has differences with the original structure because of the wall effect in thermal hydraulic characteristics. In this study, a method to analyze the wall effect was proposed based on the subchannel decomposition idea. This method could be used as a rough prediction of experimental structure selection. In the geometrical parameter optimization of typical structure for CSR1000 rod assembly, the comparison with CFD results confirmed its validity. The typical analysis of 5×5 rod array was also performed with this method and it suggested that such structure was a balance choice between the wall effect and experimental cost.

Study on Onset of Flow Instability by Genetic Neural Network

LI Jingjing, ZHOU Tao, DUAN Jun, XIAO Zejun, HUANG Yanping

2014, 35(2): 63-66.

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The trend of OFI heat flux with the system parameters is studied by the method of genetic neural network. The test result shows that the results of GNN agree well with the results of experiments. The errors fall in the limits of ±10%. By using the GNN model to predict the effect of parameters on OFI, we can find that the heat flux of OFI grows with the increasing of system pressure, the inlet subcooled temperature and the mass flow. The effect of system pressure on OFI is smaller than that of the mass flow and the inlet subcooled temperature.

Dynamic Impedance Analysis Based on an Artificial Viscoelastic Boundary Technology for Nuclear Power Engineering

LI Zhongcheng, FAN Hong

2014, 35(2): 67-70.

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It is an effective means to set artificial boundary by adopting 3D finite element technology to carry out seismic analysis of complex soil conditions. This paper is to realize the setup of artificial viscoelastic boundary as well as the exertion of equivalent load by applying Fortran to developing interface program. The availability of artificial viscoelastic boundary and wave input is verified through numerical calculation, and both the efficiency and precision are rather satisfactory. If apply this program to certain NPP seismic analysis of complex soil conditions, and compare the calculation result with that of the general program.

Flow Characteristic Analysis of Flow Induced Heat Exchanger Tube Vibration

FENG Zhipeng, ZANG Fenggang, ZHANG Yixiong, YE Xianhui

2014, 35(2): 71-75.

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A three-dimensional numerical model for heat transfer tube vibration induced by cross flow is proposed with two ways coupled approach. The characteristics of flow field are investigated based on the numerical model. The results show that, when Ur≤2, lift force coefficient increases as Ur increases, while drag force coefficient decreases first and then increases. When 2＜Ur＜9, the Locked In phenomenon occurs and drag force reaches the peak value prior to that of lift force. As Ur≥9, lift force is tending to zero. The drag force of flexible tube is about 2~2.5 times of that of the fixed tube, and the lift force is 6 times. Vibration makes the pressure distribution and wake vortex mode change. The transverse space between vortexes is amplified obviously compared to the case of fixed tube. There is only 2S mode existing in the present study.

Ultrasonic Examination for Safe End to Nozzle Dissimilar Metal Welds of Steam Generator

WANG Zhuowei, YU Jingsheng, WANG Jianjun

2014, 35(2): 76-78.

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The safe-end weld of steam generator is narrow seam weld with dissimilar metal, the filling material is nickel alloy 152/182（material 690）. The interior structure is of great anisotropic, and fake signal may occur during the defect detection by ultrasonic wave and the error for defect location may be increased. Stratified inspection by ultrasonic transducers with different angle and focus is a practical method which is verified by the real inspection while the linear indication in the inside surface besides the interior flaws are detected.

Optimization of TOFD Detection for Thick-Walled Nuclear Pressure Vessel with Inner Surface Defect

DAI Zhen, WANG Xin, JING Shangqian, LI Wei, HAO Xiaojun, WANG Lei

2014, 35(2): 79-83.

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The 20° probe is proposed in order to optimize the TOFD detection. Acoustic wave characteristics, detection signal and image feature of 20° probe are analyzed. The optimal frequency is determined by comparing 2.25MHz and 5MHz. Through TOFD detection of artificial inner surface defect, the optimum detections are demonstrated, which are fit for detecting thick-walled nuclear pressure vessel with inner surface defect. The experimental results show that:（a） for thick-walled nuclear pressure vessel with inner surface detect, 20° probe and 5MHz frequency are optimization.（b） By this optimization detection, artificial inner surface defect with height of 2mm can be effectively distinguished in nuclear pressure vessel with thickness of 120mm.

Nuclear Safety Review on Exceeding Allowable Flaw Treatment of Reactor Pressure Vessel

ZHANG Lin, CHU Qibao, FANG Yonggang

2014, 35(2): 84-85,97.

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Based on the nuclear safety review on the defect beyond tolerance of Unit 2 reactor pressure vessel of Tianwan NPP and Unit 4 reactor pressure vessel of Qinshan Phase II NPP, to the standards for the safety evaluation of the fracture mechanics of RPV defects beyond tolerance, the defect characterization, the fracture toughness, the selection of residual stress and the thermal shock under load are discussed, in order to be useful for the follow-up of the nuclear safety review.

Study on Mixed Analysis Method for Fatigue Analysis of Oblique Safety Injection Nozzle on Main Piping

LU Xifeng, ZHANG Yixiong, AI Honglei, WANG Xinjun, HE Feng

2014, 35(2): 86-89.

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The simplified analysis method and the detailed analysis method were used for the fatigue analysis of the nozzle on the main piping. Because the structure of the oblique safety injection nozzle is complex and some more severe transients are subjected. The results obtained are more penalized and cannot be validate when the simplified analysis method used for the fatigue analysis. It will be little conservative when the detailed analysis method used, but it is more complex and time-consuming and boring labor. To reduce the conservatism and save time, the mixed analysis method which combining the simplified analysis method with the detailed analysis method is used for the fatigue analysis. The heat transfer parameters between the fluid and the structure which used for analysis were obtained by heat transfer property experiment. The results show that the mixed analysis which heat transfer property is considered can reduce the conservatism effectively, and the mixed analysis method is a more effective and practical method used for the fatigue analysis of the oblique safety injection nozzle.

Design of Diverse Actuation System in Nuclear Power Plant

XIAO Peng, LIU Hongchun, ZHOU Jixiang, GUAN Zhonghua

2014, 35(2): 90-93.

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This paper analyzes the current condition of defense in depth and diversity for Fujian Fuqing nuclear power plant, and illustrates the design flow, design criterion, system structure and the design point. The result of accident analysis for Fuqing nuclear power plant unit 1 and 2 indicates that DAS is capable of mitigating the result of SWCCF in digital safety instrumentation and control system and improving the safety of the nuclear power plant, and it is an effective method to cope with SWCCF in digital safety instrumentation and control system.

Research on Compound Controller in Nuclear Power Plants

ZHU Hao, WEI Gang, ZHAI Chunrong, CHEN Qiunan

2014, 35(2): 94-97.

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The reactor power control system is the core of the reactor control system in the nuclear power plant. Taking an experimental reactor as the research object, the compound control structure including the state feedback and the PID controller to control the nuclear reactor power are adopted. The existence of the state feedback array can obviously improve the dynamic characteristics of the power system, restraining the controlling quantity of the system effectively; PID controller plays the role of fine-tuning the system. The simulation results show that the method is correct and effective.

Discussion of Important Safety Requirements for New Nuclear Power Plants

ZHANG Lin, JIA Xiang, YAN Tianwen, LI Wenhong, LI Chun

2014, 35(2): 98-100.

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This paper presents the analysis of several important safety requirements and improvement direction. Technical view of security goals on site safety evaluation, internal and external events fortification, serious accident prevention and mitigation, as well as the core, containment system and instrument control system design and engineering optimization, and etc are indicated. It will be useful for new plant design, construction and safety improvement.

Key Points of Nuclear Safety Review on Thread Remedying with Threaded Inserts

SUN Haitao, CHANG Meng, WANG Baoxiang, GAO Chen, LING Ligong, MA Ruoqun, JIA Panpan

2014, 35(2): 101-104.

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Fitting threaded inserts is a common means of remedying of internal threads deterioration founded in flange of nuclear equipments, by which defective internal threads can be repaired rapidly. Moreover, the bearing capability and anti-fatigue capability can be improved. During nuclear safety reviewing on non-conformance of threads deterioration, more attention need to be paid to the conformance with standards, material of inserts, implementing program, service analysis, inspection measures and feedback in order to ensure the reliability and quality of threads remedy.

Modeling and Simulation on Main Cooling System of China Experimental Fast Reactor

CHEN Wuxing, XIA Genglei, PENG Minjun

2014, 35(2): 105-109.

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The China Experimental Fast Reactor（CEFR） is a sodium-cooled fast reactor, whose operation characteristics of primary and secondary loops exert significant influence on reactor safe operation. The primary and secondary loops main cooling system models and the steam generator（SG） model are established by the system modeling tool JTopmeret. The codes can be edited to calculate the system parameters as the flow, pressure, temperature and others at any point of this system. Under the conditions of steady state and transient state, the errors among the simulation value and design value of system main parameters are less than 2%, which satisfies the precision requirement of system simulation.

Analysis on Modification of Containment Filtration and Exhaust System of CPR1000

ZHAO Xin, YE Ziqing, CHEN Li, SHEN Renmin

2014, 35(2): 110-113.

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Based on the practical engineering of the CPR1000 Nuclear Power Plant, this paper presents the layout improvement of the containment filtration and exhaust system. System independence of single unit and the aseismic performance is optimized by doubling the systems and improving the arrangement, which promotes the safety performance and reliability after severe accidents. The effect of the improvement and the relevant solutions are also considered. The summary and feasible analysis will offer a reference for the system design of following nuclear power station.

Research on Suppressant Characteristics of Suppression Containment

QUAN Biao, JIANG Xiaoyu, CHEN Zhihui, FAN Kai, WANG Liang, TANG Bin, YANG Junming

2014, 35(2): 114-117.

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Through the containment pressure and temperature response analysis with the level of 100 MWe reactor after a postulated accident such as LOCA, the suppressant characteristics of suppression containment have been researched in this paper. Passive spraying can not effectively decrease the pressure of containment due to the excessive releasing enthalpy in the phase of blow-down, but the suppression containment system can effectively decrease the pressure during a short term of LOCA. Through the analysis of the important parameters, such as the total volume of suppression pool, the proportion of gas to water and flow area of pipelines, it can effectively restrain the pressure of containment in the phase of blow-down, these parameters will influence suppressant effect: firstly the total volume of suppression pool is the most important factor; secondly the proportion of gas to water and the total flow area of pipelines should be optimal.

Comparative Study on Two Different Seal surface Structure for Reactor Pressure Vessel Sealing Behavior

CHEN Jun, XIONG Guangming, DENG Xiaoyun

2014, 35(2): 118-120.

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The seal surface structure is very important to reactor pressure vessel（RPV） sealing behavior. In this paper, two 3-D RPV sealing analysis finite models have been established with different seal surface structures, in order to study the influence of two structures. The separation of RPV upper and lower flanges, bolt loads and etc. are obtained, which are used to evaluate the sealing behavior of the RPV. Meanwhile, the comparative analysis of safety margin of two seal surface structural had been done, which provides the theoretical basis for RPV seal structure design optimization.

Numerical Analysis on Complete Characteristics of LOCA Reactor Coolant Pump

FU Qiang, LONG Yun, ZHU Rongsheng, YUAN Shouqi, XI Yi

2014, 35(2): 121-126.

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For the study on complete characteristics of LOCA reactor coolant pump, three-dimensional modeling for the pump internal flow channel by pro/E, based on the Reynolds-averaged Navier-Stokes equations with the RNG k-ε turbulence model and SIMPLEC algorithm, using the computational fluid dynamics software CFX to conducts the numerical simulation calculation for complete characteristics of reactor coolant pump, complete characteristic curves have been analyzed, and reasons of the hump phenomenon in small flow conditions are explained. The result shows that: numerical simulation method is feasible to calculate complete characteristics of reactor coolant pump. The impeller torque is mainly from the blade, and the torque of the front and rear shroud has little effect on the impeller torque. The torque and axial force of blade have similarity trend with Q-H curve, and they are closely related.

Study on Step Motion Characteristics of Control Rod Drive Mechanism

LIU Pengliang, ZHOU Jianming, LU: Yonghong

2014, 35(2): 127-130,172.

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The step motion of control rod drive mechanism is accomplished through the lift armature of claw assembly hoisting and falling alternation movement in the vertical direction. In the base of the control rod drive mechanism step motion decomposition, the dynamic calculation model of coupled magnetic-electric-mechanical movement is established by finite element calculation method. The relations between magnetic force, electric current, displacement and time are studied in the process of control rod drive mechanism stepping motion. The movement characteristic parameters as hoisting time, falling time, electric current and magnetic force are obtained. The results can be applied to the time sequence design of control rod drive mechanism.

Design Improvement of CPR1000 Steam Generator Supports

REN Hongbing, XIE Honghu, ZHOU Peng

2014, 35(2): 131-133.

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The design basis of earthquake resistance for CPR1000 reactor coolant system is only 0.2g, and the design abundance is low. This paper described an improved steam generator support structure with adding sway struts and connecting clevis, and eliminating gaps among supports. The stress analysis indicates that this support structure can enhances the earthquake resistance of the system from 0.2g to 0.3g, and the requirements of HAF102 are completely met.

Research on Dimension Calculation of CPR1000 Safety Injection System High-pressure Restriction Orifices

ZHAO Xin, TAN Haibo, LIU Bo, WANG Jianghong, ZHAI Bajing

2014, 35(2): 134-136.

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This paper starts from the basic principle of throttle orifice pipe. Based on the flow capacity test data of safety injection system of the 4th unit in Ling’ao Nuclear Power Plant Phase II, and by the calculation with classical fluid mechanics formulation, the theoretical results of the size for the throttle orifice are obtained, which are similar with the actual installation data, therefore, this test result can provide feasible plan for the size calculation of high pressure throttle orifice in the CPR1000 safety injection system.

Investigation on Performance of Gas Separators in Gas Removal System for MSR

ZHANG Nana, YAN Changqi, SUN Licheng, LIU Wei, LI Hua

2014, 35(2): 137-140.

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Three gas separators with different structural parameters were designed and tested with air and water as the working fluid to find the optimal structure. Visualized method was used in the experiments, with a high speed video camera employed to record the separation process, aiming to find the effect of the structural parameters. It is showed that the inlet angle, outlet angle, length and number of the vanes have significant influence on the separation process. For designing a gas separator, the outlet angle of swirl vanes to the axial direction should not be over 45° and at least 5 swirl vanes with the length exceeding 45 millimeters are required.

Design and Research on a New Circular Plate-Shell Heat Exchanger

WANG Jiazhuo, YAN Changqi, DING Ming, CHEN Zheyu, SHI Shuai

2014, 35(2): 141-145.

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A new circular plate-shell heat exchanger was designed to solve the problem that the plate heat exchanger can not withstand the high pressure.This heat exchanger has good prospects for industrial applications because it can bear high pressure and is smaller and lighter.A lot of experimental research has been done to study its heat transfer and pressure drop characteristics when it was used as lubricating-oil cooler.The experiments showed that the circular plate-shell heat exchanger had excellent heat transfer performance and it is appropriate for high viscosity fluid’s heat-transfer process.

Theory Resolution of Extraction-Insertion Force for Irradiation Capsule

XU Xiao, JIN Ting, YANG Jingchao

2014, 35(2): 146-149.

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With the theory mechanics method, the Spring-Ball model is simply equivalent to the system of irradiation capsule structure. Moving on the contact surface, the balance of the Spring-Ball is related with extraction force, surplus size, frication, spring stiffness and shape of the contact surface. The theory resolution of the Extraction-Insertion force is solved by making the balance equations of this Spring-Ball model. Comparing with the Finite Element Analysis, the analysis with theory resolution is better at the optimizing structure design, the sensitivity of the related factors.

Cracking Analysis of 304L Stainless Steel Panel on Refueling Pool in Nuclear Power Plants

CAO Feng, FANG Jiang, TANG Shiyan, ZHANG Ting, GE Lianwei, DING Youyuan

2014, 35(2): 150-153.

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The reason of cracking of 304L stainless steel covered on the refueling pool was found out through the analyses of macrostructure, microstructure, chemical composition, mechanical properties, morphology of fracture and chemical composition of corrosion products. The results show that it was typical stress corrosion cracking of 304L stainless steel induced by Cl-.The main source of Cl-is the Cl emulsion copolymerization used in the course of the execution of concrete. It was suggested that clear the concrete with Cl emulsion copolymerization thoroughly during subsequent maintenance of the stainless steel panel, reinforce the water quality control and concrete addition management during similar concrete executions.

Study on ¹³³Xe Stripping Time of Nuclear Power Station during Shutdown

LAN Lijun, LIU Jie, CHEN Yide, TANG Shaohua, ZHANG Yujia

2014, 35(2): 154-156.

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This paper studies the mathematical relationship between stripping time and stripping factor, other parameters, and builds the calculation model of ¹³³Xe stripping time under station shutdown. The result shows that the stripping time reduces to the minimum value by stripping factor rising under the definite primary coolant and stripping rate. The relation equation to select a suitable stripping factor is obtained, and the analysis relation equation about the ¹³³Xe source and the stripping rate are built.

Analysis and Countermeasures for Effects of Grid Frequency Change on PWR Unit Operation

DING Weidong

2014, 35(2): 157-160.

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In many analyzed events by WANO, operators relied solely on the reactor protection system to protect the reactor core without taking pre-emptive conservative actions. Especially during this off-normal grid conditions which was beyond the procedure guidance, these conditions may be further exacerbated by a reactor operator’s inexperience. In this paper, the potential adverse effects of the grid frequency change on PWR units operation is analyzed, the appropriate contingency plans are proposed so that the reactor operators can respond correctly in time when this off-normal grid condition occurs, to avoid causing reactor scrams to protect the reactor core, and the operators can conservatively monitor the PWR units operation, and a prudent margin to the plant safety limitations is always maintained.

Study on Characteristics of U-Mo/Al-Si Interaction Layers of Dispersion Fuel Plates

LIU Lijian, YIN Changgeng, CHEN Jiangang, SUN Zhanglong, LIU Yunming

2014, 35(2): 161-165.

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Abstract:
In this paper, we analyzed the characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates. The results show that the interaction layers（IL） are with irregular morphology and uneven thickness, and are mainly formed in the internal micro cracks of the dispersion fuel particles or at the interface between the particles and the substrates. The diffusion mechanism of U-Mo/Al-Si is the vacancy diffusion, Al and Si are migrating elements, and the diffusion reaction is that Al and Si diffuse to U-Mo alloy. Inside the interaction layers, the Al content keeps constant basically, but the Si content gradually increases with the substrate-fuel direction, and the maximum content of Si appears interaction layers near the U-Mo side. Adding about 5wt% Si into Al matrix can restrain the diffusion reaction, and improve the performance of dispersion fuel plates finally.

Research on HFETR Ageing Management and HFETR Ageing Management Database System Development

ZHANG Xiaomei, JIA Yaqing, LIU Peng, CHEN Qibing, LI Ziyan, ZHANG Ying

2014, 35(2): 166-169.

Abstract(10) PDF(0)

Abstract:
HFETR, as the multipurpose research reactor, was designed and built by China independently, and the research on ageing management of HFETR was started in 2005 after HFETR had been operated for 25 years. The research is mainly about ageing management methods, ageing management object and etc. According to the achievement of the research, HFETR ageing management database system has been developed.

Analysis on NPP Chemistry Management System

FAN Weiwei

2014, 35(2): 170-172.

Abstract(5) PDF(0)

Abstract:
The major function, action and flow chart of chemical management system for a nuclear power plant is introduced in detail, and its application in the world is illustrated. Referring to years of experience in the development of the chemical data management system in nuclear power plants, this paper discusses how to achieve the purpose of rapid identification and diagnosis by the application of artificial intelligence technology and chemical anomaly data, to meet the needs of plant operation, achieve the aim of safe, reliable and economical operation of nuclear power plants.

2014 Vol. 35, No. 2