核动力工程

Study of the Transform Method of Multi-Group Nuclear Cross Section Covariance Matrix

Wang Dongyong, Hao Chen, Zhao Qiang, Wu Zongpei, Wu Hongchun, Li Fu

2016, 37(2): 1-6. doi: 10.13832/j.jnpe.2016.02.0001

Abstract(28) PDF(0)

Abstract:
Through the analysis of the characteristics of the nuclear cross section covariance matrix, the method of transforming the nuclear cross section covariance matrix in multi-group form into the users’ group structures has been studied and a general code T-COCCO based on the method has been developed. The nuclear cross section covariance matrix built-in SCALE6.1 is applied and the nuclear cross section covariance matrix of some nuclides such as ²³⁵U, ²³⁸U, and ²³⁹Pu in 44 groups structures has been expended or collapsed into different users defined group structures, such as 33 groups, 47 groups and 70 groups. At the same time, the corresponding covariance matrix has been produced by using NJOY for comparison and the matrix eigenvalue and rank has been also studied for verification. The analysis indicates that the method studied in this paper is reasonable and T-COCCO code is convenient friendly and efficient for users to gain the desired multi-group nuclear cross section covariance matrix, which can be applied to uncertainty and sensitivity analysis of nuclear data and physical calculation.

Research on Fuel Rod Multi-Physics Numerical Modeling and Program Development

Fei Jingran, Si Shengyi, Chen Qichang

2016, 37(2): 7-12. doi: 10.13832/j.jnpe.2016.02.0007

Abstract(32) PDF(0)

Abstract:
In this paper, the concept of Numerical Fuel Rod is proposed, and the technology of multi-physical modeling is studied. The thermal conduction module, the mechanical module（both based on finite element method） and the neutron transport module（based on method of characteristics） have been programmed and tested. The calculation results show that the modules are effective, which lays a solid foundation to achieve the multi-physical-coupling calculation and simulate the fuel rod behavior in the reactor.

Physical Characteristics Study of Power Ramp Test Irradiation Rig with ³He Loop

Zhang Liang, Qiu Liqing, Deng Caiyu, Tong Mingyan

2016, 37(2): 13-18. doi: 10.13832/j.jnpe.2016.02.0013

Abstract(43) PDF(1)

Abstract:
Power Ramp Test（PRT） of a fuel element is generally conducted with a PRT irradiation rig in a research reactor, to study the fuel behaviour and to verify its safety under power transient. Neutronics characteristics of the PRT irradiation rig within a typical HFETR（High Flux Engineering Test Reactor） core and the heat generation rates of the rig’s components are calculated with MCNP code in this paper. The range of the test fuel rod power is obtained with a coupled Nuclear-Thermal-Hydraulic calculation method which combines MCNP and CFX code.The results show that the ³He gas layer influences the neutron field intensely by reducing the thermal neutron current into the layer and decreasing the neutron flux in and near the irradiation rig. Changing the density of ³He gas can vary the PRT and its periphery neutron field, consequently changing the test fuel rod power effectively. Power of the fuel pellet in the test rod increases monotonically along with the ³He gas pressure reducing, and its calculation results have good agreement with the curve fitting by a natural logarithm function.

Study on Management Strategy for In-Core Irradiated Fuel Assembly Developed by Self-Reliance

Liao Hongkuan, Li Ligang, Liao Zejun, Yu Yingrui, Deng Zhixin, Jiao Yongjun, Wu Lei

2016, 37(2): 19-22. doi: 10.13832/j.jnpe.2016.02.0019

Abstract(22) PDF(0)

Abstract:
In order to execute the irradiation plan of the N36 characteristic assembly, CF2 fuel assembly and CF3 fuel assembly which are developed by CNNC with independent intelligent property right, and meet the requirement of irradiation, this paper has analyzed the fuel management strategy from cycle 9 to cycle 12 of Qingshan-Ⅱ NPP unit 2. Comprehensively considering the economy, security and the irradiation requirements of the assemblies, 5 fuel management strategies are presented, in which 3 assemblies achieve their irradiation goals while operates economically and safely.

Experimental Study on Hydraulic Characteristics of Tube Support Plate in Steam Generator

Wen Bo, Li Yong, Xiong Ting, Zan Yuanfeng, Yan Xiao, Zhuo Wenbin, Li Pengzhou

2016, 37(2): 23-26. doi: 10.13832/j.jnpe.2016.02.0023

Abstract(26) PDF(0)

Abstract:
An experimental study on hydraulic characteristics of Tube Support Plate（TSP） with different holes shaped like broached-trefoil and broached-quatrefoil was carried out under the conditions of single-phase and two-phase flow. The results show that the single-phase local resistance coefficient of TSP decreases with the increasing of Reynolds number, and becomes constant afterwards. The bigger the radius of the chamfer of the hole or the ratio of the flow area in the TSP and the tube bundle, the smaller the single-phase local resistance coefficient of TSP. The local pressure drop is slightly larger than the calculation values by the correlations of Chisholm and Lin Zhonghu, and coincides well with the results by Computational Fluid Dynamics（CFD）.

Experimental Study on Convective Heat Transfer of Supercritical R134a in Vertical Circular Tubes

Chen Jiayue, Xiong Zhenqin, Xiao Yao, Gu Hanyang

2016, 37(2): 27-31. doi: 10.13832/j.jnpe.2016.02.0027

Abstract(22) PDF(0)

Abstract:
In order to investigate the convective heat transfer at supercritical pressures, experimental research was conducted on heated Freon R134 a flowing upward through an I.D 25 mm vertical circular tube under supercritical pressure condition. The experimental data covers a wide range of conditions. The pressure is at 4.5MPa and 4.7MPa. The mass flow flux varies from 400 to 700 kg/（m²·s） and the heat flux ranges from 30 to 60 k W/m². Both enhanced and deteriorated heat transfer were analyzed and the parametric sensitivity was also carried out. The heat transfer performance was evidently enhanced near the pseudo-critical point. The deteriorated heat transfer appeared at lower mass flow velocity or at higher heat flux under a certain ratio of q/G=0.06 k J/kg. At a mass flux of 500 kg/（m²·s） two types of deteriorated heat transfer were observed in the experiment: the first type appeared at the near entrance region of the tube and existed within different range of fluid inlet temperature; The second type appeared at any section inside the tube than entrance but only within a certain enthalpy range. Heat transfer can be enhanced by increasing mass flow velocity, decreasing heat flux or decreasing pressure, while the variance of the heat transfer deterioration is opposite.

Study on Mechanism of Radiation Heat Exchange for High Temperature Pebble Beds

Wu Hao, Gui Nan, Yang Xingtuan, Tu Jiyuan, Jiang Shengyao

2016, 37(2): 32-37. doi: 10.13832/j.jnpe.2016.02.0032

Abstract(24) PDF(0)

Abstract:
In this paper, the mechanism of radiation heat exchange for high temperature pebble beds is studied by view factor, and the analytical solution for view factor between two unit spheres is derived. Compared with various numerical methods for view factors, Monte Carlo Method（MCM） is selected with high accuracy and good parallel performance. As blocked by other particles, the view factor between particles approximately is reduced to zero at distance of double sphere diameter. With spatial local partition by Voronoi tessellation, for body center cube（BCC）, face center cube（FCC） and random packing all particles interacted by thermal radiation are included among double Voronoi spheres, about 60~80 spheres. Within acceptable range many particles with extremely low view factor are neglected, numerical radiation model is greatly simplified.

Scaling Analysis of the Transient from Forced Circulation to Natural Circulation

Shi Yan, Yang Fuming, Li Yuquan

2016, 37(2): 38-42. doi: 10.13832/j.jnpe.2016.02.0038

Abstract(36) PDF(0)

Abstract:
Integral thermal hydraulics test plays a key role in the safety verification of PWR. The scaling analysis of the circulation characteristics in the primary loop provides the theoretical basis for the integral test facility design. Based on the two-phase drift flow model, the governing equations of forced circulation and natural circulation were established. The initial conditions were applied to nondimensionalize the governing equations resulting in the similarity criterions of transient from the forced circulation to the natural circulation for the integral test facility. The experimental method which can simulate the prototype main pump coast-down and satisfy the similarity requirements of the transient between circulations was proposed.

Analysis of Downcomer Channel Thermal-Hydraulic Phenomena of PCCS Outside Cooling

Li Le, Li Cheng, Zhang Yajun, Yi Xiongying

2016, 37(2): 43-47. doi: 10.13832/j.jnpe.2016.02.0043

Abstract(26) PDF(0)

Abstract:
The structures simplified from the outside cooling of the passive containment cooling system（PCCS）, was numerically simulated, with the aim to study its effects on the residual heat removal during nuclear accidents. Characteristics of the PCCS outside cooling only with natural forces driving were obtained, based on which the coupled heat transfers among the downcomer channel, the baffle plate and the riser channel were discussed. Finally, the width of the downcomer channel on the PCCS outside cooling was analyzed, and it indicated that the downcomer channel should be considered for long vertical section.

Experimental Study on Effect of Buoyancy and Flow Acceleration on Heat Transfer of Supercritical CO₂

Liu Guangxu, Huang Yanping, Wang Junfeng, Zan Yuanfeng, Lang Xuemei

2016, 37(2): 48-51. doi: 10.13832/j.jnpe.2016.02.0048

Abstract(23) PDF(0)

Abstract:
Experiments were conducted in a natural circulation loop to investigate the effect of buoyancy and flow acceleration on the heat transfer of supercritical CO₂. A new heat transfer correlation is developed based on the experimental data. Results show that the buoyancy is the main influencing factor for the heat transfer deterioration in natural circulation condition. The new correlation could predict the heat transfer of supercritical CO₂ well.

Criticality Excursion Analysis

Yang Junyun, Xiao Gang, Ying Yangjun

2016, 37(2): 52-55. doi: 10.13832/j.jnpe.2016.02.0052

Abstract(34) PDF(1)

Abstract:
Ramp Feed experiments performed on TRACY are simulated with the code Cri SAA which is used for the criticality excursion analysis. The transient characteristics of open criticality excursions are studied. The predictive results are shown. The statistics characteristics, including time and energy to first peak, power of first peak and the total energy, are obtained by simulations. The effects of the reactivity insertion rate and the uranium concentration on the criticality excursion are analyzed. Fuchs model are improved to be suitable for open criticality excursion in this paper. It is helpful to understand the open excursion.

Numerical Investigation of Heat Transfer Characteristics of Supercritical Carbon Dioxide in Double D-Shape Channel

Liu Shenghui, Huang Yanping, Liu Guangxu, Wang Junfeng, Zhao Dawei, Zang Jinguang, Zan Yuanfeng, Lang Xuemei, Xu Jianjun

2016, 37(2): 56-59. doi: 10.13832/j.jnpe.2016.02.0056

Abstract(28) PDF(1)

Abstract:
This paper presents the results of numerical investigation of the heat transfer characteristics of the supercritical carbon dioxide in double D-shape channel. The capability and thermal efficiency（ε） are affected obviously by the mass flow rate of the working fluid, which were studied closely here. The study shows that as the mass flow rate is slow down in the hot channel, the capability decreases and the thermal efficiency raises up. Heat transfer in the heat exchanger is driven by the temperature difference, so when the hot mass flow rate slowing down, the heat flux becomes weak and the wall temperature drops down. This phenomenon is different from the fixed heat flux heat transfer.

Experimental Study on Distribution Regions of Flow instability and Boiling Crisis in Parallel Twin Rectangular Channels under High Pressure

Tang Yu, Chen Bingde, Xiong Wanyu, Huang Yanping, Xu Jianjun

2016, 37(2): 60-64. doi: 10.13832/j.jnpe.2016.02.0060

Abstract(25) PDF(0)

Abstract:
An experimental investigation has been made on flow instability and boiling crisis in a 50 mm×2 mm parallel rectangular channels at the pressure of 12~15 MPa. In the experiment, it has been found that the thermal parameter range of boiling crisis was extended, while the flow instability zone was reduced with the increase of pressure and mass velocity. And the boiling crisis curve and flow instability boundary line was drawn in N_pch-N_sub pictures to analyze the effect mechanism of the thermal parameters on distribution regions.

Investigation of TVD Scheme Applied Into One Dimensional Flow Problems

Liu Wei, Yan Xiao

2016, 37(2): 65-69. doi: 10.13832/j.jnpe.2016.02.0065

Abstract(29) PDF(1)

Abstract:
TVD is a kind of advanced numerical convection discretization scheme, which has good numerical stability and calculation accuracy. It can avoid the numerical oscillation and false diffusion, so it has the potential to be applied in the transient thermal-hydraulic analysis with wave characteristics. In this paper, a program was developed based on the TVD scheme and a one-dimensional problem is used as an example to validate the program with different cases. This paper proves the validity of TVD scheme.

A Simulation of Hysteresis Phenomenon during Cladding Transient Oxidation

He Xiaoqiang, Yu Hongxing, Jiang Guangming, Dang Gaojian, Wu Dan, Zhang Yu

2016, 37(2): 70-73. doi: 10.13832/j.jnpe.2016.02.0070

Abstract(22) PDF(0)

Abstract:
The hysteresis phenomenon related to the coexistence of monoclinic-tetragonal phases of zirconia has been analyzed by a developed cladding oxidation model based on diffusion equations. Main findings are summarized as follows: the model can well simulate the hysteresis phenomenon during cladding transient oxidation; when the heating or cooling rates are high, the hysteresis phenomenon is significant; and the model can give a much better prediction of the hypothetical loss-of-coolant accident（LOCA） oxidation experiments in Oak Ridge National Laboratory（ORNL） than the parabolic rate correlation.

Detecting Technique for Uranium Distribution Uniformity of Dispersion Uranium-Zirconium Fuel Core Based on Dual Energy Gama-Ray Transmission

Luo Jiandong, Du Yu, Tang Yueming, Xu Guiping, Wang Xuequan, Zhang Xiaochuan

2016, 37(2): 74-76. doi: 10.13832/j.jnpe.2016.02.0074

Abstract(25) PDF(0)

Abstract:
Dual energy gama-rays transmission method is a new method for the detection of uranium distribution uniformity of dispersion fuel core. The 136 ke V and 264 ke V high energy of Se-75 radiation source is chosen for the test. HPGe is detected by Gama-ray detector. Under the energy of 264 ke V,U and Zr’s gamma ray attenuation coefficient, which is far greater than the coefficient at the energy of 136 ke V, the content and uniformity of U and Zr in dispersion fuel core were determined. Plexiglass bottles contain Uranium and Zirconium powder are used as criterion specimen for the attenuation measurement and iterative method is used for solve the equation. The results present that the relative errors of detecting system is less than 5%, well meet the requirements.

Effects of Statistical Variations of Geometric Parameters of Coated Fuel Particles on the Failure Fraction

Li Rong, Liu Bing, Tang Chunhe

2016, 37(2): 77-81. doi: 10.13832/j.jnpe.2016.02.0077

Abstract(28) PDF(0)

Abstract:
The silicon carbide coating layer of the coated fuel particles was regarded as the bearing shell in the model. Considering the statistical distribution of the geometric parameters of coated layers into the pressure vessel failure model, the effects of the statistical variations of geometric parameters on the failure fraction were studied with the method of the Monte Carlo simulation. The results indicated that the failure fraction of the coated fuel particles was related to the particle geometric parameters such as kernel radius, thickness of porous layer, thickness of the inner pyrocarbon layer and thickness of silicon carbide, and kernel radius and thickness of porous layer were the main effect factors.

Thermal Property Analysis of Accident Tolerance Fuel Pellet

Xu Duoting, Liu Tong, Ren Qisen, Huang Heng, Wu Hailong

2016, 37(2): 82-86. doi: 10.13832/j.jnpe.2016.02.0082

Abstract(26) PDF(0)

Abstract:
In this paper, CFX has been used to simulate the thermal behaviors of UO₂-Mo, UO₂-SiC, UO₂-BeO, U-Si, FCM and current UO₂ fuel pellet under normal condition and accident condition, then evaluation has been made based on the simulation results. It turns out that FCM fuel has advantages due to its ability of thermal conductivity and resistance.

Quantitative Characterization of Micro-defect in Zircaloy Oxide Films

Long Chongsheng, Wei Tianguo, Chen Hongsheng, Xiao Hongxing, Zhao Yi

2016, 37(2): 87-90. doi: 10.13832/j.jnpe.2016.02.0087

Abstract(27) PDF(0)

Abstract:
Basing on ion migration in xide films, a original method named IMM（Ion Migration Method） has been established to characterize the micro-defects in oxide films quantitatively. Several kinds of zircaloy oxide films formed at different corrosion conditions, were measured by IMM. On the basis of the results, it is concluded that IMM can distinguish the micro-defect difference between the different samples and the micro-defects evolution during the corrosion sensitively. After IMM has been further improved, it will be a effective method to analyse the microstructure of oxide films.

Protective Control and Simulation Research on Nuclear Power Plant Coolant Average Temperature

Qian Hong, Fang Zhenlu, Jin Weixiao, Zhou Lei, Zheng Pengyuan

2016, 37(2): 91-96. doi: 10.13832/j.jnpe.2016.02.0091

Abstract(22) PDF(0)

Abstract:
The dynamic matrix predictive controller was designed based on the nuclear power plant coolant average temperature and using the non parametric model of the coolant average temperature. Connecting the nuclear power simulation model by the application of the OPC technology, the coolant average temperature control simulation research platform is constructed; In this simulation studies, with the control of dynamic matrix predictive controller and by a small range of load fluctuation, the tracking curve of the coolant average temperature and R bar curve are obtained. The calculation of the axial power distribution in the core deviation proved that the control algorithm can meet the control performance and guarantee the safety and rationality of the reactor power distribution.

Study on Level 2 PSA Source Terms Analysis of 1000 MW PWR

Chen Qiaoyan, Yang Zhiyi, Zhou Tao, Li Hanchen

2016, 37(2): 97-101. doi: 10.13832/j.jnpe.2016.02.0097

Abstract(26) PDF(0)

Abstract:
In order to resolve the key problem in the severe accident source term calculation technique, the release category of 1000 MWe PWRs and the release fraction of typical accident sequence are calculated by integral code. How to choose the typical scenario is a key problem. The sensitivity analysis and comparison result show that the typical scenario should come from the probability analysis rather than the experience judgement. Other key techniques which have influence on the PSA result have been studied. In order to get reasonable results, the model of the release pass and the water purify effect must be simulated in detail.

Study on Emergency Planning Zone for Small Modular Reactors

Chen Wenjun, Jiang Shengyao, Liang Manchun

2016, 37(2): 102-105. doi: 10.13832/j.jnpe.2016.02.0102

Abstract(28) PDF(2)

Abstract:
The method of emergency planning for large nuclear reactors can not be applied to small nuclear reactors directly. The division of the emergency planning zone should give full consideration to the expected dose and the dose that may be affected by the public, and to determine the scope of the emergency plan by comparing the obtained dose data with the prescribed intervention level. Take the plume emergency planning zone outside which the expected dose not more than 10 m Sv as example, calculate the release of the source item level requirements under the constraints of the emergency planning zone by using MACCS, and give a detailed description of emergency plan zone of division and the calculation process.

Design of CRDM Digital Current Adjustment Equipment Based on Transfer Phase Trigger and Close Loop Control

Li Guoyong, Zheng Gao, Xu Mingzhou, Jin Yuan

2016, 37(2): 106-110. doi: 10.13832/j.jnpe.2016.02.0106

Abstract(39) PDF(0)

Abstract:
The paper describes the design of CRDM digital current adjustment equipment which is based on the transfer phase trigger theory. The current adjustment equipment is ameliorated through PID close loop control, the real time monitoring of time sequence current. For meeting the requirement of DAS, the current rack-off function is designed. The test result shows that the equipment meets the operation requirement of CRDM, and some technical points are optimized.

Research on Time-Varying Characteristics for Nuclear Power Valve Failure Probability

Guo Haikuan, Cai Qi, Zhao Xinwen, Zhang Yongfa

2016, 37(2): 111-115. doi: 10.13832/j.jnpe.2016.02.0111

Abstract(29) PDF(0)

Abstract:
Nuclear power valve has the characteristics of high reliability and long lifetime, and its failure data have obvious small sample problem. The nuclear power valve failure are caused by loss factors of impact, shaking, abrading and corrosion. The failure probability has time trend, and the p will raise（decrease） with the increasing of time. The Jeffreys prior model of invariable p can not reflect the time trend of p very well, so it can not analyze the time-varying characteristics of p. This paper built the Generalized Linear Model（GLM） for valves failure data having Binomial distribution, analyzed the time trend of p inspected the model through posterior predictive distribution, and assessed the model ability of replicating observed data by graph inspection and Bayesian chi-square. The results showed that GLM had well fit index and was more propitious to assess the failure probability p of valve.

Independent Development of a New-Style Spent Fuel Storage Rack

Mo Huaisen, Yuan Chengyu, Tan Jingyao

2016, 37(2): 116-121. doi: 10.13832/j.jnpe.2016.02.0116

Abstract(22) PDF(0)

Abstract:
A new type of spent fuel storage rack is developed independently. Based on the structure of the spent fuel storage rack, this paper demonstrates the security and reliability in terms of the criticality safety, thermal hydraulic and seismic analysis. In order to verify the design rationality and manufacturing feasibility of the spent fuel storage rack, a prototype is manufactured, and the supporting leg adjustment test and the mock-up fuel insertion test are carried out. The prototype manufacturing and test showed that the design is reasonable, the manufacture is feasible, and the new type of spent fuel storage rack can be fabricated in mass production.

Study on Characteristics and Sensitive Factors of PCS Air Flow Path Resistance

Pan Xinxin, Xiang Wenjuan, Song Chunjing

2016, 37(2): 122-126. doi: 10.13832/j.jnpe.2016.02.0122

Abstract(24) PDF(0)

Abstract:
The air flow path resistances of the passive containment cooling system（PCS） in advanced nuclear power plants are analyzed with computational fluid dynamics（CFD） method which is verified and validated by tests. The results show that the resistance factors vary little as the flow velocity increases, while the main losses generated at air inlets, turning vane and inner annular. The flow resistances can be depressed by optimizing the turning vane geometry. Increasing the space of inner annular can reduce the resistance factor. Considering the heat transfer enhancement, the space ratio 1∶3 between inner annular and outer annular is appropriate.

Stress Calculated by Two Models of Large Storage Tank

Huang Wen, Tan Tiancai, Ma Janzhong

2016, 37(2): 127-128. doi: 10.13832/j.jnpe.2016.02.0127

Abstract(35) PDF(0)

Abstract:
Two methods for mutual verification calculation of simplified model and fluid solid coupling model are used in the stress analysis of the large storage tank. The calculation for pulsed liquid quality M₀, the height of gravity of pulsed liquid H₀, the height of gravity of liquid convection H₁, and the equivalent stiffness k₁ of spring simulating slosh frequency of liquid convection is based on HOUSNER theory and as the input parameters of finite element model. FLUID80 elements are used in the fluid solid coupling model to model the fluid and SHELL181 elements to model the storage tank. Fluid solid coupling model considering the structure damp is used to calculate the response of the tank under seismic time history.

Airtightness Test Technology for Reactor Compartment

Lin Xiaoling, Tian Yanjie

2016, 37(2): 129-131. doi: 10.13832/j.jnpe.2016.02.0129

Abstract(30) PDF(0)

Abstract:
Two different methods to test the reactor compartment airtightness have been set up, which are the pressure reduction and the indicator leakage. Measurement and analysis ways in the two methods have been separately presented according to their principle bases. Based on the theory analysis, SF₆ has been verified to act as the indicator in the indicator leakage test.

Development and Application of Routing Verification Program in Nuclear Piping Layout

Liu Qin, Chen Xinghua

2016, 37(2): 132-135. doi: 10.13832/j.jnpe.2016.02.0132

Abstract(26) PDF(0)

Abstract:
The work of piping layout in the nuclear islands of nuclear power stations includes the pipe and the support designing, which shall satisfy the requirementss of system, vibration, weight and flexibility. This paper have introduced the routing verification program during the basic design in TAISHAN project. The program is implemented in the whole design process, and the result shows that this program can effectively verify the mechanics of the layout of process pipelines, ensure the design quality, reduce the iterative work and improve the efficiency.

Analysis and Operation of Tuning Amount of Boron in PWR NNP’s Transient Test Condition

Liu Daoguang, Yu Hang, Luan Zhenhua, Liu Peng

2016, 37(2): 136-140. doi: 10.13832/j.jnpe.2016.02.0136

Abstract(32) PDF(0)

Abstract:
Dramatic changes in the status of the CPR1000 unit’s transient test process will introduce a lot of negative reactivity in a short time. So that the circuit is not easy to maintain a stable core state, and the test process requires accurate adjustment by primary coolant boron concentration to compensate for the reactivity changes. Through research G/R rods changes, power shifts, xenon poisoning variation caused by changes in the reactivity during the test, and based on the current value of the boron concentration and the differential, we analyzed the calculated amount of change in boron concentration and the amount of boron dilution during the test process and provided appropriate controlling strategies of the unit for operating personnel based on the results. Debugging process and test results show that boron dilution accurate analysis can significantly reduce the amount of the unit to control risk, improve the quality of the unit, access to good economic returns.

Research on Atmospheric Steam Dump System Control Commands Oscillations and Valve Failure of CPR1000 Nuclear Units

Peng Haicheng, Yang Zhong, Yang Jianhui, Ouyang Hui, Li Jin

2016, 37(2): 141-142. doi: 10.13832/j.jnpe.2016.02.0141

Abstract(31) PDF(0)

Abstract:
For CPR1000 nuclear units atmospheric steam dump system（GCT-a）, one event happened was mainly caused by the sustained oscillation of the valve control command which led to the valve malfunction eventually. By optimizing the PID parameter of control loop and testing the valve parameters with Flow- scanner 6000, and also considering the operation data, it is demonstrated that the valve failure is caused by the design defect on the cage contributing to the control command oscillation and damage on sealing surface, and the resolution is proposed.

CFD Study on Cavitation Abrasion for Axletree of Main Pump of Tianwan NPP

Zeng Xiaokang, Zhou Huihui, Xiong Wanyu

2016, 37(2): 143-146. doi: 10.13832/j.jnpe.2016.02.0143

Abstract(34) PDF(0)

Abstract:
Water is used as the lubricant for the axletree by Tianwan nuclear power plant. The heat generation by rate is on the high side, and serious abrasion is found on the axletree. The paper studies the characteristics of the flow and pressure in the axletree room by computed fluid dynamics（CFD） method for the abrasion of the axletree. Analysis shows that the reason for the abrasion and high heat generation rate of the axletree is cavitation; and the cavitation is caused by the flow acceleration effect of local area.

Feasibility Study on Flow Test Procedure for AP1000 Accumulator Discharge Lines

Kong Xiangwei, Qiu Fengxiang, Na Hongwei

2016, 37(2): 147-150. doi: 10.13832/j.jnpe.2016.02.0147

Abstract(28) PDF(0)

Abstract:
In this paper, the process level and injection flow rate during the accumulator discharge lines flow test are calculated with an iterative method, and the effectiveness of the test settings is evaluated against the results. It shows that the initial parameters chosen by the procedure are appropriate, and they can maintain the two check valves on the injection lines fully open.After the motor-operated valve is fully open, the injection time is still long enough for obtaining data to calculate the real flow resistance of the lines.

Typical Accident Analysis of Supercritical Water-Cooled Reactor

Liu Liang, Zhou Tao, Chen Jie, Fang Xiaolu, Chen Juan, Wei Xiaoyan, Xia Bangyang

2016, 37(2): 151-155. doi: 10.13832/j.jnpe.2016.02.0151

Abstract(28) PDF(0)

Abstract:
CSR1000 was selected as the research object. A code named SCAC-CSR1000 has been developed based on the SCAC code. The reliability of the code was verified by comparing the results of SCAC-CSR1000 and SCTRAN. Then the safety analysis was carried out. Five events were selected, that are partial loss of reactor coolant flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure and loss of feed water heating. By the numerical analysis, it was found that the MCST does not exceed 1260℃, and meets the design safety requirements. The 2nd MCST are higher than the 1st MCST. The isolation of main steam line has the less safety margin.

Mechanism Study on Passive Siphon Break of a Low-Pressure System

Zheng Yiyun, Huang Shanfang, Guo Xiaoyu, Wang Ningbo, Pang Tianfeng, Jiang Bin, Qin Xiaobin, Xia Shaopeng, Li Min

2016, 37(2): 156-159. doi: 10.13832/j.jnpe.2016.02.0156

Abstract(34) PDF(0)

Abstract:
Siphon break has great influence on the Loss of Coolant-Accident（LOCA） for the nuclear safety. Experimental researches were carried out to study the siphon breaking process in a passive air-water two-phase flow loop. Flow processes were observed and important parameters were measured including pressure drops, water flowrate, and water level. Flow pattern in downward flow pipe is found to be bubbly flow, slug flow and falling film flow in turn. Experimental results show that the passive siphon breaking system could significantly increase pressure drop and reduce water flowrate. Therefore, under the LOCA condition, the goal to reduce the exposing risk of the reactor core can be realized by the siphon breaking effect.

Experimental Study on Flow Characteristics of Large Subcooled Water in In-Containment Refueling Water Storage Tank Based on Scaled Model of AP1000 ADS under Depressurization Condition

Wu Guanghao, Lu Daogang, Wang Zhongyi, Zhang Yuhao, Fu Xiaoliang, Yang Yanhua

2016, 37(2): 160-164. doi: 10.13832/j.jnpe.2016.02.0160

Abstract(28) PDF(0)

Abstract:
Based on the design of AP1000 power plant, a fully scaled experimental mock-up was built to simulate the high-temperature and high-pressure steam spraying, steam condensation under depressurization condition through Automatic Depressurized System（ADS）. The experiment recorded the flow characteristics and 3D temperature and velocity distribution around sprayer in IRWST by dot matrix of thermocouples and Particle Image Velocimetry Particle Image Velocimetry（PIV）. By studying the thermal stratification and natural circulation in the IRWST under depressurization condition when undercooling water is large, an optimization scheme was proposed to improve the efficiency of cooling water in IRWST.

CFD Analysis on Key Components of Mixing Vane Grids—Vane Angle/Dimple Shape

Yang Baowen, Han Bin, Zhang Hui, Shan Jianqiang

2016, 37(2): 165-170. doi: 10.13832/j.jnpe.2016.02.0165

Abstract(35) PDF(0)

Abstract:
As the most important structure to enhance fuel assembly and thermal hydraulic performance, spacer grids are of great importance to improve the economic efficiency and the safety of nuclear reactors. The numerical simulation and experimental study of the mixing vane grids design have drawn much attention in the past years. Key components, such as mixing vane and dimple, in spacer grids have a large impact on the mechanical and thermal hydraulic performance. In this paper, the simulation pressure drop value was compared to the experimental data to provide initial validation of the CFD method. Then the effects of mixing vane angle and dimple shape on the flow field downstream of spacer grids were analyzed based on the validated CFD model, meanwhile the code-to-code comparison between CFX and STAR-CD simulation results was also made to give an initial evaluation of CFD calculation.

Experimental Study on Thermal Fragmentation of Melt

Peng Cheng, Tong Lili, Cao Xuewu, Yan Xiao

2016, 37(2): 171-174. doi: 10.13832/j.jnpe.2016.02.0171

Abstract(24) PDF(0)

Abstract:
Experimental study on the thermal fragmentation of melt, including Sn, Pb and Sn-Pb alloy, has been carried out based on the SSFT（Small-Scale Fragmentation Tests） facility. The effects of the melt properties, release distance, initial melt temperature, and coolant temperature on the thermal fragmentation have been studied. By analyzing the debris characteristics and the distribution, several specific thermal fragmentation mechanisms have been given, corresponding to different parameters. Finally, a partition map of thermal fragmentation mechanisms has been drawn based on the previous work.

Demonstration of Simulation Analysis Program PCCSAP-3D for Passive Containment Cooling System（PCCS）

Wang Yan, Yang Yanning, Zhang Yaoli, Zhou Zhiwei

2016, 37(2): 175-179. doi: 10.13832/j.jnpe.2016.02.0175

Abstract(27) PDF(0)

Abstract:
PCCSAP-3D is a dedicated analysis code developed for the Westinghouse AP1000 PCCS. In this paper, the models on the AP1000 reactor system are described. The PCCS performances during the postulated design basis accidents（LOCA and MSLB） simulated and analyzed from PCCSAP-3D are compared with that from the WGOTHIC developed by Westinghouse. The results show good agreements, which indicate that the contain-ment pressure during the postulated design basis accident can be limited effectively under the design value by the PCCS performance and also the capacity of the PCCSAP-3D for the analysis and the evaluation on the PCCS performance is validated preliminarily.

Numerical Simulation of Double-Sensor Conductivity Probe Measurement Mechanism for Two-Phase Flow

Liu Hang, Pan Liangming, Deng Jiajia, Yuan Dewen, Huang Yanping

2016, 37(2): 180-184. doi: 10.13832/j.jnpe.2016.02.0180

Abstract(35) PDF(0)

Abstract:
Based on the model of VOF（Volume of Fluid） and UDS（User Defined Scalar）, this paper builds the simulation model of using conductivity probe measuring two-phase flow parameters. The process of using double-sensor probe measuring two phase flow was simulated. The electric field distribution was obtained when the probe pierced the bubbles. The results show that the probe piercing the bubbles cause the distinct change of the distribution of current and voltage. As the process of simulation is not influenced by noise signal, bubble shape variation and signal response delay, the ideal signal is obtained such as the square wave signal of current and voltage. The simulation results truly reflect the basic process of the measurement by using the double-sensor probe for gas liquid two-phase flow.

2016 Vol. 37, No. 2