核动力工程

Transient Analysis on Overpower and Loss of Heat Sink of A Small Natural Circulation Lead-Cooled Fast Reactor

Shi Kangli, Zhang Xilin, Chen Hongli

2017, 38(4): 1-5. doi: 10.13832/j.jnpe.2017.04.0001

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Abstract:
Small natural circulation lead-cooled fast reactor has a good property of passive safety and economic. The primary cooling system model was established, based on the concept of a 100 MW_th small natural circulation lead-cooled fast reactor（SNCLFR-100）. With RELAP5 code, initial state and transient conditions were analyzed, including the protected transient of overpower and loss of heat sink, and the unprotected transient of overpower and loss of heat sink. The results show that during the protected transient of overpower and loss of heat sink process, the reactor is safe with the help of reactor trip protection system. As for the unprotected one, the reactor shuts itself down on account of the strong negative feedback of reactivity in 500 s, and the temperatures of coolant, cladding and fuel pellet are below the safety limits. The simulations of the two transient could verity the inherent safety of this new reactor.

Experiment Study on Thermal Erosion of CF2 Series Fuel Assemblies

Tian Xuelian, Nie Changhua, Yu Qinglin, Rong Xiaohong, Xu Zhangzhe, Li Shuo, Chen Xungang

2017, 38(4): 6-10. doi: 10.13832/j.jnpe.2017.04.0006

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Abstract:
A full size and truncated（used in small modular reactors） CF2 fuel assembly were adopted in the hot rushing test. The experiment was carried out in the condition of high temperature and pressure, where the components of the driving line were simulated as 1:1. The reliability and anti-abrasion of fuel assembly was tested. The dropping performance data were obtained for the two sizes of fuel assembly. Moreover, the method of rushing test was also discussed in this paper. It provided an important experimental basis for the optimization design, safety assessment and application of domestic advanced fuel assembly.

Turbulence Model Evaluation in Circular Pipe of Supercritical Water Based on Statistical Method

Leng Jie, Zang Jinguang, Yan Xiao, Li Yongliang, Huang Yanping

2017, 38(4): 11-15. doi: 10.13832/j.jnpe.2017.04.0011

Abstract(26) PDF(0)

Abstract:
Up to now, the traditional turbulence model evaluation in supercritical water still has some uncertainties to some degree. The basic reason is the limit of model itself which makes the model depend on the calculation parameters. This paper focuses on the turbulence model evaluation based on the batch process techniques of CFD tools through comparing with the experimental data base. Some typical turbulence models were analyzed against the experiments. The performance of each model was evaluated based on the statistical method. The dependencies of the performance on operation conditions were discussed.

CFD Simulation and Synergy Field Analysis of Two Typical Mixing Vanes of Space Grad

Dai Xudong, Chen Weihong, Yang Lixin

2017, 38(4): 16-21. doi: 10.13832/j.jnpe.2017.04.0016

Abstract(17) PDF(1)

Abstract:
Based on the validation of CFD models with Hollway experiment results, two groups of CFD models about 5×5 rod bundle with separate or split mixing vane grids were respectively setup and simulated. The field synergy principle was used to discuss the mechanism of heat transfer enhancement by mixing vane according to the CFD results. The study shows that the field synergy principle is feasible to explain the mechanism of heat transfer enhancement in fuel assembly and the heat transfer enhancement can be predicted by the synergy angle distribution. If the pressure loss is ignored, the performance of the split mixing vane is superior to the separate mixing vane based on the enhanced heat transfer. Increasing the blending angle of separate mixing vane does not significantly enhance the heat transfer in the rod bundle, and even prevent the heat transfer at a large blending angle.

Transient Study on Characteristics of Power Generation by FHR Coupling with Air Brayton Turbine Cycle

Ruan Jian, Zou Yang, Li Minghai, Zhou Bo, Zhu Guifeng, Xu Hongjie

2017, 38(4): 22-26. doi: 10.13832/j.jnpe.2017.04.0022

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Abstract:
Based on the characteristics of power generation and thermodynamic cycle of the mark-1 fluoride salt cooling high temperature reactor system, the thermoelectric conversion efficiency of the FHRs coupled with air Brayton cycle system is studied. By studying the gas turbine calculation method, the effect the air flow rate and inlet air temperature on FHRs and the variation laws of FHR efficiency under non-rated conditions are analyzed. The results show that the thermal efficiency is up to 42.6%, and the efficiency decreases with the increasing of the air flow rate. Less than 5% change of air flow rate can keep the efficiency changing within 2%, and different reheating strategies of the system can be selected if the change of the flow rate exceeds the scope. In addition, the thermal efficiency of the system decreases with the increasing of the air inlet temperature, and the change of the system efficiency is less than 1%.

Effect of Damping Parameters on Seismic Analysis of Nuclear Reactor Coolant Loop

Wang Qing, Fang Yonggang, Chu Qibao, Lu Yan, Xu Yu, Li Hailong, Wen Jing

2017, 38(4): 27-30. doi: 10.13832/j.jnpe.2017.04.0027

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Abstract:
The Rayleigh damping parameters and its effect on the seismic analysis for the nuclear power plant reactor coolant loop are studied in this paper. Using the same model and other input parameters, three groups of Rayleigh damping parameters are used to calculate the seismic response of the reactor coolant loop. Typical support loads in the results are selected and analyzed. With the increasing of the natural frequency and effective mass of high frequency of the reactor coolant loop model, greater value of damping parameter has great effect on the calculation results.

Shaking Table Test for Evaluation of Seismic Behavior of Nuclear Power Plants on Non-Rock Site

Li Xiaojun, Wang Xiaohui, HE Qiumei, Liu Aiwen

2017, 38(4): 31-35. doi: 10.13832/j.jnpe.2017.04.0031

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Abstract:
In order to study the suitability and seismic response characteristics of the nuclear power plants on non-rock site, taking CAP1400 nuclear power plant（NPP） structure as an example, the shaking table test considering soil-structure interaction（SSI） was completed. The test study indicates that the site can amplify the peak acceleration which presents in all different directions, and the structure can influence the site response spectra in low frequency. Cracks appear on the model site when inputting less than safe shut-down earthquake（SSE）, and the contact surface between the structure and the soil separates when inputting safe shut-down earthquake（SSE）. At the end of the test, the surface cracks on the model site were connected. The model structure had no crack. The damage of the model system was due to the instability of the foundation. It was obvious that the seismic response of the NPP was affected by site conditions, thus the site conditions and the SSIanalysis should be taken into account in the seismic response analysis of the NPP.

Study on Control Rod Device Mechanism Impacting on Missile Shield Plate

YE Xianhui, ZhenG Bin, Jiang Naibin, Wu Wanjun, Xu Pei

2017, 38(4): 36-38. doi: 10.13832/j.jnpe.2017.04.0036

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Abstract:
Using the fluid-solid interaction arithmetic of ANSYS/LS-DYNA, the dynamic analysis for control rod device mechanism（CRDM） impacting on missile shield plate is performed, and the intensity of plate is evaluated. The shock analysis shows the impact at the rim of the plate is more dangerous than that at the center. When the rim of the plate to be evaluated is under impact, its stress intensity exceeds the yield strength and some part of the material is plastic. When the strain intensity method is adopted, the result shows the computed strain still has quite amount of margin. The plate will not be penetrated and can remain its integrity, preventing control rod missile to impact other components.

Comparative Study of Welding Procedure Qualification Standard between NB/T 20002 and ASME-Ⅸ in Nuclear Power Industry

Wang Yuxin, Li Zhe, Dong An, Wang Heng

2017, 38(4): 39-42. doi: 10.13832/j.jnpe.2017.04.0039

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Abstract:
This paper studies the differences on structure and content of the welding procedure qualification standard NB/T 20002 and ASME-Ⅸ for the G-three nuclear power station ACP1000 and AP1000. This paper selects butt-welding with Shielded Metal Arc Welding（SMAW）, which is the most widely used welding process, analyzes the differences between NB/T 20002 and ASME-Ⅸ, and makes a comparison study on testing and acceptance level of welding procedure qualification, and finally gives a evaluation conclusion.

Effect of Different Designs for Loss of LBA on Core Damage Frequency

Yang Jian, BEi Junjuan

2017, 38(4): 43-46. doi: 10.13832/j.jnpe.2017.04.0043

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Abstract:
Direct current power LBA performs important safety function. There are two different designs for loss of LBA in domestic M310 units. This paper analyzes the core damage frequency of loss of LBA by using the probabilistic method, and evaluates the safety level of different designs. In terms of core damage risk, the design which unlocks the generator-transformer shutdown protect is more reasonable.

Improvement of AMS Calibration Equipment in Radiation Protection Area

Li Xiaoning, Ren Xueming, Zhou Bing, Chen Haiyan, Zhang Kai

2017, 38(4): 47-50. doi: 10.13832/j.jnpe.2017.04.0047

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There is a strong radioactive source in the EPR’s AMS calibration equipment, but the shielding design has a defect, which will increase the irradiation dose and unplanned irradiation risk. The equipment was redesigned to decrease the dose and the risk. Besides, the effect evaluation was implemented based on MCNP software calculation. Finally, the optimization of radiation protection is realized. The design theory and the calculation method are introduced in the paper.

Analysis and Application of Load Change Rate Algorithm For CPR1000 Nuclear Power Plant

Lan Bing, MenG Qingjun, Yang Jingli, Cai Yawei

2017, 38(4): 51-55. doi: 10.13832/j.jnpe.2017.04.0051

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Abstract:
According to the requirement of the controlling processing for CPR1000 nuclear power plant, the drain valve should be opened in 2~10 s, and could resist 1~2s interference. Generally, the differential algorithm or iteration algorithm can be used for the change rate calculation in DCS system, but they cannot meet the above requirement. Based on the simulation and analysis of these two algorithm, an optimized sliding average iteration algorithm for the calculation of unit load is presented. This algorithm combines the disturbance-rejection characteristics of sliding average algorithm and the change rate calculation function of cycle iteration algorithm, and meets the certain requirement. This algorithm has been applied in the DCS process control of CPR1000 several nuclear power plant.

Numerical Computation Verification of Erosion Behaviours of Reactor Cavity with High Gas-content Concrete in Homogeneous Melt Pool Configuration

MA Jian, Yan Xiao, Zan Yuanfeng, Zhuo Wenbin

2017, 38(4): 56-59. doi: 10.13832/j.jnpe.2017.04.0056

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Abstract:
OECD CCI-2 prototype material test was used for calculation case in this paper. By means of appropriate assumption, specific model combination and sensitive parameter optimization, MEDICIS code modeling analysis was conducted on the erosion behaviours of the reactor cavity with high gas-content limestone-common-sand concrete in homogeneous melt pool configuration. The calculation results have a satisfied agreement with the related experimental results. It was shown that the present modeling method is with fair flexibility and applicability, thus could supply the accident predictions on nuclear power plant with related modeling reference and validation basis.

Test of Air Cooling Characteristics of Control Rod Drive Mechanism for ACP100

Tian Xuelian, Rong Xiaohong, Zhuo Wenbin, Xu Zhangzhe, Wang Yunsheng, Zhang Yang, Xu Shijie, Zhang Lin, MA Xinguang

2017, 38(4): 60-63. doi: 10.13832/j.jnpe.2017.04.0060

Abstract(37) PDF(1)

Abstract:
The thermal test study focusing on air cooling performance of the control rod drive mechanism（CRDM） for ACP100 is carried out in the environment of practical reactor operation condition, which is simulated as 1:1. The balance temperature of the CRDM in different cooling air speed condition, the changing of the temperature after the cooling air broken off, and the electrical characteristics were measured. Some crucial data such as the minimum cooling velocity, the allowable time and the maximum performance without cooling were acquired. The test results provided an important reference for the cooling system design and operation of the small modular reactor ACP100.

Development of Intelligentized Accident Source Term Evaluation Code

Yu Hong, Li Lan, Zhu Jianping, HE Fan

2017, 38(4): 64-69. doi: 10.13832/j.jnpe.2017.04.0064

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The accident source term evaluation code can only evaluate single accident and output the evaluation result for given release routes, and requires complex pretreatment and special input card. To resolve these problems, BP-ASTE code which can simulate isotope radioactivity in each containment and isotope radioactivity release rate from containment to joined containment following accident evolution is developed based on accident sequence and matter and energy release. The example of SGTR accident is given to illuminate the establishing and implement of BP-ASTE code.

Experimental Study of Passive Injection System Operation Characteristics under Different DVI Break Size Conditions

Huang Zhigang, PenG Chuanxin, Zhang Yan, Li Yang, Zan Yuanfeng, Zhuo Wenbin

2017, 38(4): 70-75. doi: 10.13832/j.jnpe.2017.04.0070

Abstract(31) PDF(2)

Abstract:
In this paper, the passive injection system operation characteristics of an integrated modular small reactor under different break sizes of direct vessel injection line（DVI） are studied by experiments. The test result shows that the bigger the safety injection pipeline break area is, the bigger the difference injection flow rate of the break side and non beak side core makeup tank system is. The accumulator injection has obvious effect on the core makeup tank injection in the case of large break area. In the early operation stage of the core makeup tank system, the operation mode is water-water cycle. After the steam flow into the pressure balance pipeline, the operation mode is steam-water cycle. The accumulator is driven by nitrogen expansion, and the injection flow rate is affected by the pressure drop rate of the system.

Thermal Optimization and Design for Irradiation Rig of Multi-Fuel Assembly

Xu Taozhong, Xiang Yuxin, Wang Hao, DenG Caiyu, Liu Shuiqing, Sun Shouhua

2017, 38(4): 76-78. doi: 10.13832/j.jnpe.2017.04.0076

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Abstract:
The numerical model of multi-fuel assembly was built up by the ANSYS/CFX code to optimize the irradiation rig of assembly, and the experiment was operated in the HFETR. The experimental data showed the measuring data could represent the real irradiation data after optimization of the multi-fuel assembly rig.

Design Calculation of Carbon Delay Bed

Zhu Bonan, Liu Yu, Wang Yiwei

2017, 38(4): 79-83. doi: 10.13832/j.jnpe.2017.04.0079

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Based on the third generation nuclear plant, the radioactive noble gas from core fission and the model of the gas movement in the waste gas system are analyzed, focusing on the delay state analysis of the isotopes of Kr and Xe creeping through the activated carbon. This paper gives a calculation method to determine the carbon mass and the system holdup time by annual radioactive release. The dimensions are determined based on the flow velocity recommended in ANSI55.4. By comparing the delay bed dimensions parameters of a certain nuclear corporation, the results by this method are consistent with them, which proved that the calculation method and the assumption in this paper are reasonable.

Investigation on Methodology of Coastdown Analysis and Design for Reactor Coolant Pump

Zhong Yun, Liu Yi, Zhou Wenxia, XiA Di

2017, 38(4): 84-88. doi: 10.13832/j.jnpe.2017.04.0084

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Abstract:
Methodology of coastdown analysis for reactor coolant pump（RCP） was investigated with consideration of combined effect of loop fluid inertia and pump inertia. It has been verified that the calculated coastdown flow fits with the measured data better when considering the influence of loop fluid inertia. A coastdown performance oriented design method considering the influence of loop fluid inertia was developed based on coastdown analysis methodology investigation, which can increase the pump inertia flexibility without violation of nuclear safety. And the methodology of coastdown analysis and design can applied to sealless RCP when using overall efficiency without electric loss as input.

Analysis of Rigging Lifting for AP1000 Containment Vessel Module by Finite Element Method

Li Tuo

2017, 38(4): 89-93. doi: 10.13832/j.jnpe.2017.04.0089

Abstract(25) PDF(1)

Abstract:
Containment vessel of AP1000 is built by modules, and the deformation and stress should be strictly limited when module is lifted. During the AP1000 project, the containment vessel module is lifted by lifting beam. A new lifting method called rigging lifting was proposed in this paper. The module lifting was modeled in ANSYS and analyzed by finite element method, the effect of lifting lug location and rigging number on deformation and stress was studied. The maximum deformation and maximum equivalent stress of two lifting methods was compared, and rigging lifting method was proved to be feasible. The rigging lifting method can reduce the maximum deformation and maximum equivalent stress of containment modules, improving the quality of containment vessel and reducing the construction cost.

Kinematic Verticality Adjustment Technology for Refueling Machine Inner Mast

Shen Yabo, Li Tao, Guo Keke

2017, 38(4): 94-96. doi: 10.13832/j.jnpe.2017.04.0094

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RFM（refueling machine） is the main component which used to perform core loading/offloading in nuclear power plants. The kinematic verticality of RFM inner mast will directly affect the safety of fuel operation. By analyzing of RFM commissioning in a nuclear power plant, it is proved that the kinematic verticality of RFM inner mast can be accurately measured by a laser device. By the means of improving the mast installation verticality and the inner-outer mast concentricity, and accurately adjusting the inner-outer mast guide roller gap, the RFM inner mast kinematic verticality can be sufficiently improved, which will reduce the occurrence of fuel operation accidents.

Study on Availability Factor Model Promotion in Nuclear Power Plants

Chen Wen, Yang Xiaohu, Jiang Hong

2017, 38(4): 97-100. doi: 10.13832/j.jnpe.2017.03.0097

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The availability factor is a very important indicator to measure the design and operation performance of nuclear power units. It is also the economic performance indicator of WANO（The World Association of Nuclear Operators） to measure the world’s nuclear power plants. To improve the availability factor of nuclear power units is the focus and research direction of the industry. In this paper, on the basis of many years of good practices in CPR1000 nuclear power station, the availability model of nuclear power units is established, and the methods and ways to improve the design availability factor are provided, to optimize the design of new nuclear power plants and to provide the reference for the availability factor index evaluation in follow-up nuclear power plants.

Development of a γ-Ray Monitoring System Based on Local Area Network

Li Mingfu, Jin Zhangjiang, Yuan Yonggang

2017, 38(4): 101-103. doi: 10.13832/j.jnpe.2017.04.0101

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The network radiation detection system is designed based on the controller local area network（CAN） bus. This system takes Geiger-Miller counting tube（referred to as G-M tube） as a γ-ray detector, and the reliability of the communication mode of CAN bus is realized by the data transmission. The system uses modular design, and is with strong expansibility, which can realize the online real-time monitoring of the gamma radiation dose.

Cause Analysis and Suggested Solutions for Reactor Coolant Pump Motor Oil Leakage

Jiang Hong, Zhou Jing, Jiang Xiaomao, HE Jingsong, DenG Xiao, MAO Yuanfan

2017, 38(4): 104-107. doi: 10.13832/j.jnpe.2017.04.0104

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In this paper, cause analysis of reactor coolant pump motor oil leakage is performed, and suggested solutions are presented. Based on the CFD analysis result, supplementary baffle and seal can reduce the fan effect on the lower guide bearing caused by flywheel, and the supplementary condensate device and oil recycling pipe can recover the possible oil and oil mist leakage. The solutions mentioned above can reduce the oil leakage and oil mist leakage of this motor.

Research on Eddy Current Test Precision for the Position of Anti-Vibration Bars on Steam Generator Tubes

Ma Qiang, Kong Yuying, Gu Bo, Lu Ji, Wang Xiaogang

2017, 38(4): 108-111. doi: 10.13832/j.jnpe.2017.04.0108

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The position offset of the anti-vibration bar on steam generation tubes is a potential safety hazard for SG operation. In this paper, eddy current test is conducted to inspect the anti-vibration bar position. Meanwhile, the factors of the test precision were also studied. The result shows that, acquisition speed and the guide tube length have less effect on the test results. The guide tube status and the curvature radius of tube have great effects on the reliability and the stability of the test result.

Failure Analysis and Overall Maintenance of Recirculation Cooling Water System Heat Exchanger of Third Qinshan Nuclear Power Plant

Gong Daitao, Jiang Feng, Wang Xin, Chen Jifang, Hu Jianqun

2017, 38(4): 112-115. doi: 10.13832/j.jnpe.2017.04.0112

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Abstract:
The design life of the heat exchanger of the recirculation cooling water system in Third Qinshan Nuclear Power Plant was 40 years, and they are the nuclear safety related equipment. After three years of use, a large number of erosion characteristics have appeared in the tube sheet and heat exchanger tube. Through the analysis, it was found that the operation parameters of the system were higher than the design value, and the material selection, equipment selection and structural design were not suitable for the matching of the environmental characteristics of the sea water, which was the primary cause for the severe erosion of the material. And the mechanical characteristics of the local flow field were changed by the peeling off of the rubber lining, which was the direct cause for the sudden serious damage to the equipment. By completing the work such as the system transformation, tube change, tube plate repair and protection, and re-lining, the long term reliability of the equipment has been restored and improved.

Research of Test Scenario of T2 Response Time of Reactor Protection System

Xu Yong, Hu Wenyong, Wang Yan, Lin Yan, You Bing

2017, 38(4): 116-119. doi: 10.13832/j.jnpe.2017.04.0116

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The composing of the reactor protection system response time was introduced firstly. Considering the complex and difficulty of the test of RPS response time, and high requirement of test theory, two test schemes of all path converge test and all module converge test were introduced. The advantages and disadvantages of two schemes were analyzed. Through three tests during the outage in Fuqing Nuclear Power Plant, the feasibility of the test was proved.

Research and Development of Service Life and Replacement Strategy for Neutron-temperature Measurement Channel

Zhang Qi

2017, 38(4): 120-122. doi: 10.13832/j.jnpe.2017.04.0120

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The ICIS at Tianwan Nuclear Power Station is the only first in China to successfully realize reactor protection functions by linear power density and DNBR. As the kernel detectors of ICIS, NTMC plays an extremely important role. And the performance of the detector is very important to the system’s reliability. Based on the experience of TNPS, this study gives an research and development of the service life and replacement strategy for the detector.

Non-Linearity Research on Ultrasonic Inaccessible Zones for Inlet to Shell Weld of RPV

Hong Maocheng, Wu Jianrong, Yuan Shuxian, MA Guanbing, Chen Liang, Sun Jiawei, Zhu Chuanyu

2017, 38(4): 123-127. doi: 10.13832/j.jnpe.2017.04.0123

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The inlet nozzle to shell weld of PWR RPV has the characteristics of complicated geometry, while the definition of ultrasonic inspection area is different among several in-service inspection codes, and these above factors cause the inaccessible zones to show non-linearity change. In order to find the special changing rules, the parametric model of the nozzle weld was established, meanwhile the simulation method which caused by the slope feature was deeply analyzed, which help to deduce the discriminant criteria for the right-side inaccessible zones simulation. Based on the criteria, with typical parameters of the PWR RPV and ASME code, it is easier to acquire the mathematics diagram of the volume for both left-side and right-side inaccessible zones to refraction angle, with the help of lots of scattered data set by 3D simulation software. The diagram can interpret an optimization method for getting minimum inaccessible zone volume for phase array or conventional method. The research takes the advantage of increasing the chance of finding the ultrasonic defects, especially for the mathematic laws of accessibly and distribution of the phase array or conventional ultrasonic technics in same type of welds.

Measurement and Calculation of Heat Generation Rate of Stainless Steel in a Research Reactor

Si Junping, Tong Mingyan, Yang Wenhua, Zhang Liang, Nie Liangbing, Zhang Ping

2017, 38(4): 128-133. doi: 10.13832/j.jnpe.2017.04.0128

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A special calorimeter based on the static isothermal method under thermal equilibrium conditions was designed to measure the heat generation rate of the stainless steel. Besides, the distribution of the heat generation rate of the steel along with the reactor active axial region, as well as its relationship with the reactor power are further explored. Meanwhile, the MCNP code was also used to calculate the heat generation rate of the steel, and the accuracy of the MCNP code was verified by comparing the corresponding measurement and calculation values. The results show that the heat generation rate of the steel in the research reactor is closely related to the location and the reactor power, and it approximately changes along with the reactor active axial region in a truncated cosine curve. The maximum heat generation rate is located in about 50 mm below the center plane of the reactor, and it linearly increases with the increasing of the reactor power. At the range of the experiment, the values calculated by the MCNP code are averagely 18.1% higher than those acquired in the experiment, and it shows that the heat generation rate of the steel by the MCNP code has a guiding significance for the practical application in the research reactor.

Uncertainty Analysis of Manufacturer Parameters Impact on TVS-2M Fuel Rod Performance in Reactor

Jiang Xiaochuan, Song Zifan, He Kai, Yang Nirong

2017, 38(4): 134-138. doi: 10.13832/j.jnpe.2017.04.0134

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TVS-2M fuel will be deployed in the TIANWAN NPP Unit 3＆4 from the first batch. In traditional method, the predicted conservative fuel manufacturer parameters will be used to analyze the fuel rod performance under steady state condition for the first 8 fuel cycles, which includes 5 fuel rod design criteria. But this conservative evaluation method can neither reflect the realities of the real world, nor analyze the sensitivities of the manufacturer parameters. Therefore, by coupling DAKOTA and START-3 codes, GRS method is used to calculate the fuel performance and to analyze the uncertainty. After compared with the traditional method, the results reveal that the traditional method not only overestimate the fuel temperature and cladding stress, but also underestimate the axial and radial strain, and the sensitivities of cladding inner diameter and pellet outer diameter are significant for fuel rod criteria.

Microstructure of Heat-Affected Zone in a Weld Joint between Alloy 690 and Alloy 152

Nie Shuhong, Liang Zhengqiang

2017, 38(4): 139-144. doi: 10.13832/j.jnpe.2017.04.0139

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The distributions of residual strains, micro-hardness, intra-grain average misorientation（GAM）, grain boundary microstructure and metallurgical characteristics were investigated at the top, middle and root of the heat-affected zone of a weld joint between pressurized water reactor（PWR） control rod drive mechanism（CRDM） Nozzle Alloy 690 and Alloy 152 using electron backscatter diffraction（EBSD）, scanning electron microscopy（SEM） and micro-hardness tester. The results showed that the residual strain decreased from the last welding top of the weld to root, the maximum residual strain is 5.15% at the top of the last weld; the average fraction of coincident site lattice ∑3 special boundaries increased from the last welding top of the weld to root, the minimum fraction of ∑3 special boundaries is 46.6%, which is remarkably less than that 68% of the base 690 alloy. The HAZ at the top of the weld is considered to sustain the highest stress corrosion cracking（SCC） susceptibility.

Simulation Research on Effect of Processing Parameter on Hot Rolling of Fuel Plate

Yang Hongyan, Peng Xiaoming, Ding Shurong, Guo Zhen, Kong Xiangzhe

2017, 38(4): 145-148. doi: 10.13832/j.jnpe.2017.04.0145

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With the explicit dynamic finite element method, the rolling process of U10Mo/Al dispersion-type fuel plate was simulated, and the effect of rolling processing parameter on the rolling behavior and contact pressure of fuel plate was investigated. Finite element simulation results show that the rolling speed is with obvious effect on the strain rate of fuel meat. The deformation and stress condition of fuel meat would be increased with the increasing of the reduction rate. As the friction factor increasing, the spread rate reduces gradually, but the stress on fuel meat grows accordingly.

Study on Microstructure and Mechanical Properties of Zr-Sn-Nb Alloys Thin Sheet Weld

Li Shunping, Peng Qian, Yang Zhongbo, Chen Le, Li Weijun

2017, 38(4): 149-152. doi: 10.13832/j.jnpe.2017.04.0149

Abstract(17) PDF(0)

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The microstructure, microhardness, tensile properties and cyclic deformation behavior of Zr-Sn-Nb alloys thin sheet weld were studied. It reached the conclusion that FZ is composed of α' bounded by β phase boundary; HAZ is composed of α+α'. The microhardness of FZ and HAZ are higher than that of BM, and the microhardness of FZ is higher than that of HAZ. The strength of FZ is higher than that of BM, but the plasticity decreases. The cyclic deformation behavior of Zr-Sn-Nb alloys thin sheet weld is different from thin sheet at higher strain amplitude.

Study on Stress Corrosion Cracking of Welded Joint with Inconel 690 and 321 Stainless Steel

Sun Yongduo, Xiong Ru, Qiu Shaoyu, Song Yiyang, Wang Jiamei, Zhang Lefu

2017, 38(4): 153-158. doi: 10.13832/j.jnpe.2017.04.0153

Abstract(21) PDF(0)

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Stress corrosion cracking（SCC） behaviors of welded joint with inconel 690 and 321 stainless steel（including soldering seam, heat-affected zone of inconel 690 and heat-affected zone of 321 stainless steel） were studied through slow strain rate tension（SSRT） test under 100 mg/L Cl^-1without O₂ as well as 100 mg/L Cl^-1 with saturated O₂. Effects of microstructure, chlorine ion and oxygen content on SCC behaviors of the materials were analyzed through SSRT stress-displacement curves and fracture morphology. It is found that the heat-affected zone（HAZ） of inconel 690 is with low probability of SCC under 100 mg/L Cl^-1 without O₂, and showed a certain degree of SCC tendency under 100 mg/L Cl^-1 with saturated O₂, while HAZ of 321 stainless steel is with high probability of SCC under both conditions. The coarse grains of inconel 690 HAZ make against the intercoordination between grains during the plastic deformation, and reduce the strengthening and toughening of grain boundaries, and therefore, increase the SCC tendency of the HAZ.

Numerical Simulation Study on Containment Hydrogen Deflagration Risk under Severe Accident

Yang Fan, S.Kudriakov, Yu Hongxing, Deng Jian, Li Songyu, Ceng Wei, Liu Songtao

2017, 38(4): 159-162. doi: 10.13832/j.jnpe.2017.04.0159

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In this study, hydrogen risk analysis of typical Chinese nuclear power plant under LOCA scenario is performed based on CFD codes. Hydrogen deflagration potential is analyzed basing on the flame acceleration criteria. Conservative model（CREBCOM） is chosen for the combustion simulation and the burning rate is defined based on the large scale combustion experiment data. Numerical simulation showed that the containment peak pressure during the combustion process approached 7.0 bar, which would potentially compromise the integrity of the containment.

Derivation and Preliminary Verification of Algorithm for Homogenized Flux Reconstruction

Ming Pingzhou, Lu Wei, Cao Xingdi, Xia Bangyang, Liu Dong, Yu Hongxing, Sun Yufa

2017, 38(4): 163-167. doi: 10.13832/j.jnpe.2017.04.0163

Abstract(23) PDF(0)

Abstract:
The algorithm of homogenized flux reconstruction based on the plane wave expansion method is derived and programmed in the 2D（Two-dimensional） polar coordinate system. The general solution form suitable for the discrete calculation could be given by using the base expansion functions of plane wave in the diffusion equation. The preliminary programming and verification results show that the algorithm and its program implementation can achieve reasonable accuracy in the 2D lattice level.

Numerical Verification of KYLIN-Ⅱ Code Based on IAEA Plate Fuel Benchmark

Lu Wei, Yin Qiang, Chen Dingyong, Chai Xiaoming, Tu Xiaolan

2017, 38(4): 168-171. doi: 10.13832/j.jnpe.2017.04.0168

Abstract(25) PDF(0)

Abstract:
IAEA plate fuel benchmark was analyzed to verify the neutron transport function of KYLIN-Ⅱ. The numerical results show that the module of KYLIN-Ⅱ is correct and KYLIN-Ⅱ can calculate different enrichment plate fuel problem accurately.

Analysis of Natural Circulation Flow Characteristics under External Vessel Cooling of In-Vessel Retention

Yan Xiao, Hu Qiang, Huang Shanfang, Yu Junchong, Li Yang

2017, 38(4): 172-177. doi: 10.13832/j.jnpe.2017.04.0172

Abstract(26) PDF(1)

Abstract:
Based on the one-dimensional steady-state separated flow model and low-velocity subcooling model of net void generation（NVG） point, a method to study the characteristics of the natural circulation flow in the annular channel under condition of in-vessel retention has been established. The results are compared with the ULPU-V experiments which measured the natural circulation flow rates to verify the reliability and correctness of the numerical calculation model established in this work. The effects of configuration parameters and thermal-hydraulic parameters on the natural circulation for in-vessel retention of AP1000 are analyzed. The results show that the water subcooling, flow channel size, flooding levels, inlet area and loss coefficients of riser outlet have a dominant effect on the natural circulation flow rate. And the natural circulation flow rate shows various variation tendencies along with those parameters due to the driven forces and flow resistance.

Investigation on Pressure Drop Oscillation in Two-Phase Natural Circulation system

Peng Chuanxin, Zhuo Wenbin, Zan Yuanfeng, Xu Jianjun, Lu Xiaodong, Huang Yanping

2017, 38(4): 178-181. doi: 10.13832/j.jnpe.2017.04.0178

Abstract(23) PDF(0)

Abstract:
The pressure drop oscillation in two-phase natural circulation system was investigated by CATHARE code. The results show that the pressure drop oscillation was caused by the interaction between the natural circulation loop and the pressurizer in the two-phase natural circulation system. The pressure drop oscillation in two-phase natural circulation system could be avoided by limiting the fluctuating flow.

2017 Vol. 38, No. 4