核动力工程

Simulation Study on Intense Pulse Nuclear Radiation Detectors with Ferroelectric Materials

Liang Wenfeng, Wu Jian, Lu Yi, GAO Hui, Li Meng, Rong Ru

2017, 38(S1): 1-3. doi: 10.13832/j.jnpe.2017.S1.0001

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Abstract:
Based on the working principle of a ferroelectric radiation detector and the response of the read out circuit, the electrical signal as a function of material properties and the wave-forms of radiation field was developed theoretically. A code based on GEANT4 was adopted to simulate the energy depositions of the CFBR-Ⅱ leakage neutrons and gammas in three typical ferroelectric materials, including lanthanum lead zirconate titanate PLZT ceramics, lithium tantalate single crystal, and polyvinylidene fluoride films, of which the theoretical properties as nuclear detectors were obtained. To detect the radiation field of the CFBR-Ⅱ reactor, the levels of sensitivity is approximately in the range of 10^-25~10^-27 C·m^-2, and results were confirmed by experiments. Although the polyvinylidene fluoride detector has the lowest sensitivity, it still have the potential to be used for fast neutron wave-from detections due to the singnal mainly contributed by neutrons.

Design Research of Equipment for Pressing Bottom End Plug of Fuel Rods

Huang Fan, Deng Changyi, Yang Tonggao, You Yong, Chen Lei

2017, 38(S1): 4-7. doi: 10.13832/j.jnpe.2017.S1.0004

Abstract(18) PDF(1)

Abstract:
The equipment pressing the plug is a kind of important machine in which the fuel rod is pressed into the cladding tube before the electron beam or the inert gas tungsten inert gas welding（TIG welding）. The existing equipment pressing the plug can not meet the current demand. Therefore, a new type of equipment pressing the plug is designed and developed. Based on the analysis of the structure and technical requirements of the plug, the design idea of the equipment pressing the plug is determined. This paper introduces the design reason and method of the end plug feeding（storage） mechanism, the feeding end plug mechanism, the plug mechanism, the upper and the lower feeding mechanism and its subsidiary mechanism. The plug pressing process of the new designed equipment shall be able safe, stable, convenient and automatic, and without scratch after the plug is pressed.

Monte Carlo Calculation for D-T and D-D Reaction Neutrons Based on Solid Target

Qin Jianguo, Lai Caifeng, Lu Xinxin, Zhu Tonghua, YE Bangjiao, Liu Rong, Jiang Li, Wang Mei

2017, 38(S1): 8-12. doi: 10.13832/j.jnpe.2017.S1.0008

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A Monte Carlo model was modeled based on the TARGET code and titanium tritide/titanium deuteride solid target. The angular distribution for D-T and D-D reaction neutrons, differential cross sections, energy loss and mean energy of deuteron ions, mean energy and energy straggling for neutrons, depth distribution for reaction frequency, neutron flux spectra and yields were calculated based on the parameter of the actual Ti T and Ti D targets. The characteristics of the D-T and D-D neutrons were obtained and discussed. The results can be used to describe an anisotropic neutron source accurately in other Monte Carlo models, and it can also provide a reference for the selection of monochromatic neutrons and neutron yields.

Study on Second Phase Particles in N36 Alloy Claddings and Bars

Dai Xun, Wang Pengfei, Wang Ying, Cheng Zhuqing, Yang Zhongbo, Zhao Wenjin, Zhuo Hong

2017, 38(S1): 13-17. doi: 10.13832/j.jnpe.2017.S1.0013

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Second phase particles（SPPs） in N36 zirconium alloy claddings and bars was studied by SEM and TEM equipped with EDS. The result showed that SPPs in the claddings and the bars nearly have the same average diameter, but have different morphology and distribution. Zr（Nb,Fe）₂ Laves phase and a few of β-Nb phase precipitates were detected both in N36 alloy claddings and bars by diffraction and EDS analysis, but Laves phase in the bars had lower Nb content and smaller crystal lattice than that in the claddings. This study also illuminated that Nb content in α-Zr matrix for the cladding approaches to the solubility limit which appears equal to 0.3 wt~0.4 wt%, which is much lower than the usually assumed value of 0.6 wt%.

Optimization of Safety Valve Maintenance Policy for Nuclear Power Plants

Guo Song, Li Xiaozhong, Wang Yuxiang

2017, 38(S1): 18-21. doi: 10.13832/j.jnpe.2017.S1.0018

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Abstract:
At present, some nuclear power plants in China have been operating safely for more than 10 years, entering the middle stage of operation life. A lot of operational experience and data is accumulated for the system and equipment. The existing safety valve maintenance policy is based on the early design experience and the analysis assumptions of PSA（ probabilistic safety assessment）. These assumptions do not reflect the actual operational experience and feedback of the operating nuclear power plants. The reliability of these equipment is increased with the current technological developments. In this paper, the reliability-centered maintenance（RCM） method is used to optimize the maintenance policy of these safety valves.

Study on Implementation Procedure for Local Official Nuclear Emergency Response

Yu Hong, Yang Shuqi, Cheng Shisi

2017, 38(S1): 22-26. doi: 10.13832/j.jnpe.2017.S1.0022

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Abstract:
Based on the criteria, content and practical implementation of emergency plan for nuclear power plants in China, this paper analyzed the emergency response actions taken in emergencies, the characteristics of each emergency response action and the sequence in emergency response action implementation, and put forward the content and the process of the implemention procedure for local official nuclear emergency response, which designated the organization authority, personnel responsibility, material distribution and prior item of local government.

Analysis of Acoustic Vibration of Steam Generator Secondary Separator

Zhang Fengshou, Ye Xianhui, Wu Wanjun, Jiang Naibin

2017, 38(S1): 27-30. doi: 10.13832/j.jnpe.2017.S1.0027

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The finite element model of steam generator secondary separator was developed, and the analysis of the structural vibration characteristics and the aerodynamic noise of the steam dryer was completed. The frail plate part of steam dryer was found based on the vibration study. In the aerodynamic noise section, the maximum static pressure and pressure power spectral density（PSD） were included. Structural strength and fatigue characteristics were studied under the aerodynamic loads. The results demonstrated that the main frail plates are the vertical partition plate and perforated plate at the steam dryer. The maximum stress intensity appears at the perforated plate. But it is less than the limit value, and meets the requirement of RCC-M specifications.

Research on Influence of Heating Temperature and Heat Treatment on Mechanical Property and Ultrasonic Test of 316LN Stainless Steel

Li Qi, SuN Lei, Jiang Xinliang

2017, 38(S1): 31-33. doi: 10.13832/j.jnpe.2017.S1.0031

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Abstract:
By adopting the ferrite tester, the precipitation regularity of ferrite of 316 LN stainless steel with different heating temperatures and cooling rates has been researched. The results show that there is no ferrite to precipitate when the heating temperature is below 1280℃, or when the cooling rate is above 20℃/h. Based on the physical and chemical testing and combined with the heat treatment process test, the cause for appearing no bottom ware during the ultrasonic test of 316 LN stainless steel forgings and the improving method have been analyzed. The results indicate that the cause for appearing no bottom ware is the serious mixed crystal. And the difference between grain sizes can be reduced by means of heat treatment, so as to improve the detectability of ultrasonic wave. The results of the mechanical property test at ambient temperature and high temperature and the intercrystalline corrosion property test after the solution treatment with different temperatures show that the mechanical property of 316 LN stainless steel has slightly fluctuated when the grade of grain size varies in the range of 4.0 and 1.0. At this time, the intercrystalline corrosion property can meet the requirements. During the solution treatment with same temperature, the number of times of heat treatment has less influence on the grain size and the mechanical property, in order to make the intercrystalline corrosion property meet the requirements of specifications.

Research on Spent Fuel Storage Tanks for Neutron Protection Structure

Luo Xiaowei, Li Xiaoming, Guo Jun, Guo Chaoxuan

2017, 38(S1): 34-36. doi: 10.13832/j.jnpe.2017.S1.0034

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Abstract:
Boron-containing polyethylene could be selected as the neutron shielding of the spent fuel storage tank outer casing shield tank. Because of it’s sensitive to temperature, if the structural design is unreasonable, the deformation of boron-containing polyethylene could be too large for the structure stability and shielding effect when the ambient temperature difference is large or the tank temperature changes greatly for accident. This may produce gaps and cracks, who scatter the ray to environment and cause pollution. Double layer structure is used for neutron shielding out of the spent fuel storage tank. The inner layer is the main shielding layer, for which expansion joints are reserved to meet large temperature difference of environment and the accident conditions. The outer layer is a supplementary protective layer, which can effectively solve the problem of leakage at the expansion joints. The top of the shielding structure is designed open, filled with boron-containing polyethylene powder. In accident conditions, the powder will melt to fill, and the height of shielding structure could be kept to avoid radiation exposure.

Calculation and Analysis of Effective Delayed Neutron Fraction β_eff in MSR

Hu Tianliang, CAO Liangzhi, Wu Hongchun, ZHuang Kun

2017, 38(S1): 37-40. doi: 10.13832/j.jnpe.2017.S1.0037

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Abstract:
The effective delayed neutron fraction(β_eff) plays an extremely important role in the dynamics behavior of the reactors. In molten salt reactors（MSRs）, because of the adoption of the fluid fuel, the calculation of the β_eff is different with that in the traditional solid-fuel reactor. This work concentrates on the calculation of the β_eff in MSRs. The β_eff of the MOSART was calculated and the effect of external-loop transit time and inlet velocity on the β_eff is analyzed in detail in this paper. The results indicate that the increasing of the external-transit time decreases the β_eff, and the increasing of the inlet-velocity also decreases the β_eff.

Effect of Spacer Grid Model on Sub-Channel Analysis Code

Dong Siying, Liu Yang, DAN Jianqiang

2017, 38(S1): 41-44. doi: 10.13832/j.jnpe.2017.S1.0041

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The sub-channel analysis codes merely conduct the analysis on the simple spacer grid, and represent the spacer as the profile drag coefficient, which can only reflect the average effect of the fluid resistance generated by the axial position and can not accurately reflect the effect of mixing vanes on their local field. In this paper, we choose the empirical correlation appropriately and introduce the geometric dimensions of mixing vanes. The term of mixing vane resistance is expressed quantitatively in the axial and lateral momentum equations of sub-channel analysis. Therefore, we establish the distributed resistance of the mixing vanes model. The improved model is applied to sub-channel analysis code, ATHAS. The effect of different shapes, angles or arrangement of mixing vanes of 5×5 bundle assembly is calculated with and without the improved model. Then the thermal hydraulic performance is analyzed accurately, especially the flow and temperature field.

Numerical Simulation of Flow in 3×3 Rod Bundle in Vertical and Inclined Conditions

Tang Hong

2017, 38(S1): 45-48. doi: 10.13832/j.jnpe.2017.S1.0045

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The purpose of the present paper is to investigate the thermal-hydraulics of rod bundle in vertical condition and inclined condition, and the fuel assembly with 3×3 rod bundle is simulated with Reynolds-Averaged Navier-Stokes equations. The flow field is analyzed. The results show that the inclination will decrease the value of bulk velocity in gaps and the bulk velocity decreases with the increased angle of inclination in conditions of same inlet velocity. Inclination changes the distribution of temperature and the maximum temperature of the wall increases with increased angle of inclination which is not benefit for the safety of nuclear reactors.

Effects of Irradiation Damage on Electro-Magnetic Properties of Reactor Pressure Vessel Steels

Li Chengliang, Li Yuanfei, Chen Jun, Liu Feihua, SHu Guogang

2017, 38(S1): 49-53. doi: 10.13832/j.jnpe.2017.S1.0049

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The study on the variation of electrical properties and magnetic properties of reactor pressure vessel steels during its irradiation embrittlement procedure were introduced, and the current research achievements and shortcomings were discussed. Finally, it is pointed out that exploration of the potential relevance between the mechanical properties, electrical properties and magnetic properties of in-service reactor pressure vessel can result in the nondestructive evaluation techniques, which provides a new way of irradiation surveillance for reactor pressure vessels, as an important supplement to the existing conventional destructive tests program.

Irradiation Damage Effects on Microstructure of Reactor Pressure Vessel Steels

Li Chengliang, Li Yuanfei, MO Huajun, Liu Feihua, Chen Jun, SHu Guogang

2017, 38(S1): 54-57. doi: 10.13832/j.jnpe.2017.S1.0054

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Transmission electron microscopy, small angle X-Ray/Neutron scattering and positron annihilation technology in the research on the reactor pressure vessel steels microstructure’ s variation during irradiation damage procedure were introduced, the current research achievements and shortcomings were also discussed. Finally, the necessity and importance of carrying out the research on the irradiation damage effects on the domestic reactor pressure vessel steels microstructure were pointed out.

Design of Hot Cell Laboratory and Equipment for HTR Spent Spherical Fuel Element

Zhao Hongsheng, Wang Taowei, Liu Xiaoxue, Xu Gang, Chen Xiaotong, Li Ziqiang, Liu Bing

2017, 38(S1): 58-61. doi: 10.13832/j.jnpe.2017.S1.0058

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HTR spent fuel laboratory is designed and built to establish an advanced platform with hot cells and ancillary facilities for irradiated HTR spherical fuel elements and spent fuel elements. The post irradiation examination（PIE） research carried along this project aims to find out the failure and damage mechanisms of both TRISO coated fuel particles and HTR spherical elements. The hot cell laboratory mainly consists of five single hot cells, six shielded glove boxes, one conveyor line and a pneumatic sample transportation system. Apparatus and ancillary equipment are specially modified based on HTR fuel element’s microstructure to meet the requirements of remote control. The hot cell laboratory covers the functions of weighting, burnup measurement, heating test, deconsolidation test, damage rate measurement, sample preparation, macrostructure and microstructure investigation, thermal properties measurement of both post irradiated spherical FE and TRISO coated fuel particles. Results will provide significant reference data for new fuel element design.

Effect of Helium Ion Irradiation on Microstructure and Properties of Anatase TiO₂ Films

Wang Jiaheng, Wang Long, Yang Jijun, Liao Jiali, Yang Yuanyou, Liu Ning

2017, 38(S1): 62-66. doi: 10.13832/j.jnpe.2017.S1.0062

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In order to investigate the microstructure and properties changes due to helium irradiation, anatase TiO₂ film specimens were irradiated with helium ions. Irradiation damage was simulated by SRIM program. For researching the phase changes, the crystalline, the structure and morphology, and the electronic resistance and reflectance of Anatase TiO₂ films before and after helium irradiation, specimens were measured by X-ray diffraction（XRD）, Raman spectroscopy（RM）, field emission scanning electron microscopy（SEM）, atomic force microscope（AFM）, four-probe resistivity tester（FPPT） and ultraviolet-visible spectrophotometer（UV）, respectively. The results show that there is no conspicuous phase change after irradiation; crystalline and electronic resistivity is modified clearly; columnar structure is removed gradually; when the irradiation dose is constant, the smaller the ion energy is, the greater the surface roughness and the light reflectivity are; the greater dose results in the smaller light reflectivity when the energy of irradiation ions is constant.

Effects of Xenon Ion Irradiation in Nuclear Graphite

Huang Qing, Li Jianjian, Huang Hefei, YAN Long, Zhu Zhiyong

2017, 38(S1): 67-70. doi: 10.13832/j.jnpe.2017.S1.0067

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Abstract:
Surface-polished IG110 nuclear graphite samples were irradiated with xenon ions. The surface morphology and irradiation damage was characterized before and after irradiation. Irradiation at room temperature induced severe anisotropic swelling of the graphite crystallites, but the swelling did not cause intergranular cracking, which was ascribed to the well-known irradiation-induced creep of unclear graphite. The severe swelling caused the shrinkage of many pores in the unclear graphite, indicating that irradiation to a certain dose would not enhance the salt infiltration into the unclear graphite in molten salt reactors. Raman spectra showed that G peak width increased monotonically with the increasing dose, and decreased gradually with the increasing of the annealing temperature, thus showed a potential to characterize the irradiation damage in the unclear graphite.

Study on Quantitative Non-Destructive Test Method for Nuclear Fuel Rods by Three Dimensions Neutron Images

Wei Guohai, Chen Dongfeng, Han Songbo, Liu Yuntao, Wu Meimei, He Linfeng, Wang Yu

2017, 38(S1): 71-73. doi: 10.13832/j.jnpe.2017.S1.0071

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Abstract:
Neutron imaging is one kind of non-destructive testing（NDT） technique. It can be used to test the defects of pellets, confirm the enrichment of the ²³⁵U, investigate the burnable poison, fix the position and measure the hydrogen in claddings, and so on. Traditional neutron imaging technique gets two dimensions neutron images. It can not be used to test the shape of the defects, the distribution of the pellets, and the condition of the broken cladding. The three dimension neutron imaging quantitative NDT method to test the nuclear fuel rods was presented in this paper, and the experiments were carried out at China Advanced Research Reactor（CARR）and HZB. The three dimension quantification of the defects can obtain 3D image of the internal impurity and measure the parameters. The three dimensions quantification of the hydrogen in the cladding can measure the hydrogen content in almost all the positions in the cladding.

Fracture Toughness of RPV Steel Using Small Specimens after Neutron Irradiation

Lin Yun, Tong Zhenfeng, Ning Guangsheng, Yang Wen

2017, 38(S1): 74-76. doi: 10.13832/j.jnpe.2017.S1.0074

Abstract(22) PDF(0)

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The domestic A508-3 steel is a kind of low alloy ferritic steel, using for the reactor pressure vessel（RPV）.The material has significant ductile-brittle transition behavior, and after neutron irradiation, it resulting in significant irradiation embrittlement effect, reducing the material toughness, increasingthe risk of brittle fracture. To master the neutron irradiation effect on the fracture toughness of the pressure vessel steel, fracture toughness tests have been done by A508-3 0.5CT specimens, and the analyzed the neutron irradiation on A508-3 steel fracture toughness combining with the fracture data before and after neutron irradiation.

Study on Configuration of Nuclei for MnNi-Enriched Precipitation in Model Reactor Pressure Vessel Steels

Wang Dongjie, He Xinfu, Dou Yankun, Wu Shi, Yang Wen

2017, 38(S1): 77-81. doi: 10.13832/j.jnpe.2017.S1.0077

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Abstract:
It has been justified that Mn Ni-enriched precipitation（MNP） is an important contribution to the embrittlement of low-alloyed reactor pressure vessel（RPV） steels（like A508 III steel） when irradiated to high dose in like 60 to 80 years. In this paper the Molecular Statics method was used to calculate the binding energy and formation energy among solutes and defects, so as to find out the possible nuclei of MNP with the lowest energy, which might be classified into two kinds of complexes, one composed by Cu and several（≥3） vacancies and the other mixed FeMn＜110＞ dumbbell with Mn, Ni or Cu atoms trapped.

Effects of Neutron Irradiation on Mechanical Properties of 15MnTi Steel

Zheng Quan, Tong Zhenfeng, Ning Guangsheng, Zhang Zhangyi, Yang Wen, Zhong Weihua

2017, 38(S1): 82-84. doi: 10.13832/j.jnpe.2017.S1.0082

Abstract(13) PDF(0)

Abstract:
Through impact tests and uniaxial tensile tests, the effects of neutron irradiation on the tensile properties and impact performance of 15 MnTi steel are studied. Impact energy of parent metal and heat-affected zone（HAZ） of 15 MnTi steel and tension curves of 15 MnTi steel at room temperature and 300℃ before and after irradiation with neutron irradiation temperature 50℃ and cumulative fast neutron flux 1.5066×10¹⁸ cm^-2 are obtained. Results show that the neutron irradiation increases the ductile brittle transition temperature（DBTT） of the parent metal and HAZ of 15 MnTi steel, in which the parent metal increases more than HAZ, while the upper shelves of impact energy change little; yield strength and tensile strength of the parent metal of 15 MnTi steel increase after irradiation, especially the yield strength at room temperature, but the tensile properties at high temperature change little.

Failure Analysis on AFA 3G Gd Rod from Nuclear Power Plants

Wang Xin, Chu Fengmin, Bian Wei, Guo Yifan, Qian Jin, Yan Tianyu, Liu Jinping, Liang Zhengqiang

2017, 38(S1): 85-88. doi: 10.13832/j.jnpe.2017.S1.0085

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The coolant radioactivity increased during the reactor operation, and a Gd rod was suspicious to be failed. The post-irradiation examination in China Institute of Atomic Energy confirmed that both ends of the Gd rod were failed, and the hydrogen content analysis showed that the hydrogen content at the position of lower plug welding was 1720μg/g, while that was only 133μg/g at the position nearby the upper plug hole welding. That means the failure at lower plug welding was a secondary failure, and the upper failure was primary. The failure root cause of Gd rod was hole welding failure.

Study on Microstructure of Zr-2.5Nb Pressure Tube Material

Guo Lina, Han Hua, Bian Wei, Chu Fengmin, Qian Jin, Liang Zhengqiang

2017, 38(S1): 89-93. doi: 10.13832/j.jnpe.2017.S1.0089

Abstract(15) PDF(0)

Abstract:
Zr-2.5Nb alloy with hydrogen content 60 mg/g was produced by electrolytical hydrogenation. The microstructure of the Zr-2.5Nb alloy was investigated using X-ray diffraction（XRD）, optical microscope（OM）, and transmission electron microscopy（TEM）. Fractography was performed on the delayed hydride cracking（DHC） fracture surfaces using scanning electron microscopy（SEM）. The results have shown that the matrix of the Zr-2.5Nb alloy is close-packed hexagonal structure, and the grain size is larger（more than 5 mm） away from crack-tip, but smaller（approximately 1um） near the tip of the crack. The hydride with stripy feature is δ-ZrH_1.66 of face-centered cubic structure; it distributes parallel to the rolling direction. The second phase Nb particles, which have the body centered cubic, varying from 50 nm to 500 nm in size, are roundly shaped, appear quite homogeneously distributed in the inner region of the grains, but they were cluster on grain boundaries. When the preset crack direction is parallel to the rolling direction and the test temperature is 250℃, DHC fracture growth process is an intermittent process. Each step of crack propagation results in the crack extension by a distance approximately the length of the hydride, and the length of the hydride is about 20 mm.

Investigation on Behavior of Delayed Hydride Cracking in Zr-2.5Nb Pressure Tube Material

Bian Wei, Guo Lina, QiAN Jin, CHu Fengmin, Wang Huacai

2017, 38(S1): 94-98. doi: 10.13832/j.jnpe.2017.S1.0094

Abstract(12) PDF(0)

Abstract:
In this paper, the behavior of delayed hydride cracking in Zr-2.5Nb pressure tube material containing 60 μg/g hydrogen has been investigated. During the experiments, the DHC crack growth was monitored by the force fluctuation. The crack propagation rate（DHCR）and the threshold stress intensity factor（ KIH） were calculated later. Moreover, the influence of temperature and rolling direction was investigated. The results have shown that when the test temperature is 250℃, the rang of DHCR and KIH is 5.15×10^-8~15.14×10^-8 m/s and 16.55 ~18.49 MPa m^1/2, respectively. When the test temperature is 200℃, the rang of DHCR and KIH is 2.11×10^-8~2.36×10^-8 m/s and 26.22~30.89 MPa m^1/2, respectively. As the temperature decreases, the DHCR reduces while the value of KIH increases. When the preset crack direction is vertical to the rolling direction, the DHC phenomenon has not been observed.

Modeling of Evolution of Radiation-induced Frank Loops in Austenitic Stainless Steels

Wu Shi, He Xinfu, JiA Lixia, Dou Yankun, Wang Dongjie, Yang Wen

2017, 38(S1): 99-104. doi: 10.13832/j.jnpe.2017.S1.0099

Abstract(14) PDF(0)

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This work mainly focuses on the evolution mechanism of Frank loops in austenitic stainless steels under neutron irradiation by computer simulation. Molecular dynamics and mean field rate theory were employed. Firstly, a rate theory model was constructed to model the evolution of the neutron irradiation induced Frank loop in austenitic steels; Secondly, this model was parameterized by molecular dynamics simulation, and the model was validated by comparing the simulation results of dislocation loops with the experimental data in steels under electron and neutron irradiation conditions. Based on our results, the effect of survival defects under cascade collision was studied, and the survival fraction, the cluster fraction of defects and the fraction of 4-interstitials-clusters are the mainly factors for the evolution of Frank loops. The fraction of interstitial clusters which size are smaller than 4 has no effect on the Frank loop evolution under neutron irradiation.

Study on Stress in Oxide Formed on Cladding of Fuel Rod from Qinshan Phase Ⅰ NPP

Tang Qi, Wang Huacai, Fu Cheng, Liang Zhengqiang

2017, 38(S1): 105-109. doi: 10.13832/j.jnpe.2017.S1.0105

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Zirconium alloys are often used as nuclear fuel cladding in pressurized water reactors. Oxidation of these alloys is a limiting factor in the lifetime of the fuel. To extend the burn-up of the nuclear fuel requires the control of the corrosion, it is necessary to study the cladding corrosion process. The composition and the residual stress of the oxide films of the spent fuel rod from Qinshan Power Plant Phase Ⅰ were investigated by XRD method and stress measurement instrument. The results indicate that the compressive stress exists in the oxide film, and the value decreases gradually from the bottom to the top of the fuel rod. However, the stress at the air chamber suddenly increases after it gradually stabilizes at the minim value. Also, the stress has very important effect on the stability of the t-ZrO₂, and the composition of the film gradually transforms to the monoclinic phase along with the releasing of the stress caused by the development of cracks.

Raman Spectroscopy Analysis of Oxide Film Formed on Cladding of Spent Fuel Rod from Nuclear Power Plants

Wang Huacai, Tang Qi, Fu Cheng, Liang Zhengqiang

2017, 38(S1): 110-114. doi: 10.13832/j.jnpe.2017.S1.0110

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The crystal structure of the oxide film of Zr-4 alloy of different positions from the bottom of fuel rod, which comes from Qinshan nuclear power plant phase 1, is investigated by Raman spectroscopy method. The results indicate that the bottom of the rod has better corrosion resistance and more tetragonal ZrO₂, and has the black appearance which is compact. With the increasing of the distance from the bottom, the corrosion is more and more severe, the average content of tetragonal ZrO₂ in the oxide film decreases, that of monoclinic ZrO₂ increases, and tetragonal ZrO₂ transforms into monoclinic ZrO₂, and there is a gradual transition from the black and white appearance into white and porous appearance. In the cross section, from the oxide/metal interface to the surface of the oxide film, the content of tetragonal ZrO₂ reduces gradually and the monoclinic ZrO₂ increases. The result is similar to the investigations in the autoclave, i.e., the transformation from tetragonal ZrO₂ to monoclinic ZrO₂ decides the corrosion resistance of Zr-4 alloy in the reactor, and the higher the m-ZrO₂ content in the oxide film is, the higher corrosion rate of cladding, and the worse corrosion resistance property.

Hardening Mechanism Induced by Cr Precipitates: an Atomic Study

Jia Lixia, He Xinfu, Dou Yankun, Wu Shi, Wang Dongjie, Yang Wen

2017, 38(S1): 115-120. doi: 10.13832/j.jnpe.2017.S1.0115

Abstract(14) PDF(0)

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Thermal and irradiation would result in the formation of Cr enriched α’ phase in ferritic steels. The Cr precipitates can impede the mobility of dislocation, causing material hardening and embrittlement. Knowing the interaction mechanism between dislocation and Cr precipitates can help understand the irradiation hardening of ferritic steels. Using molecular dynamics（MD）, the interaction between Cr precipitates and edge dislocation（ED） in BCC-Fe system was studied; the effects of precipitate size, Cr content in precipitate and cutting position on the pinning strength of Cr precipitate were estimated. The results show that: dislocation shear the Cr precipitate leaving no defects; the pinning strength increases with the size of Cr precipitate; Cr precipitate with higher Cr content is a stronger obstacle for dislocation; it shows highest CRSS when dislocation cutting along the equator plane of Cr precipitate.

Ab-initio Calculations of Structural, Electronic and Energetic Properties for Pure α-Zr Crystal with Single Vacancy

Wen Bang, Pan Rongjian, Wu Lu, ZHang Wei, Wu Xiaoyong, HE Wen, Kharchenko Vasyl O., Kharchenko Dmitrii O.

2017, 38(S1): 121-124. doi: 10.13832/j.jnpe.2017.S1.0121

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Ab initio calculations of structural, electronic and energetic properties for pure α-Zr with different concentrations of isolated vacancy were made by using software packages Wien2 k. The lattice constant change in the pure α-Zr with different concentrations of isolated vacancy was obtained, and its electronic density, density of states, and band structure were calculated. The results show that an increase in the concentration of the isolated vacancies in the pure α-Zr leads to the decreasing of the values of the lattice constants. The electronic density of an atom which has a vacancy as a first-nearest neighbor becomes stretched in the direction of neighboring atoms, whereas the electronic density of those atoms, which have zirconium atoms, as first-nearest neighbor, remains symmetric. With the decreasing of the concentration of the isolated vacancy inside the unit cell, the height of main peaks of total density of states increases, and the band structures become more complicated.

Assessment of Irradiation Embrittlement of Domestic RPV Material

Sun Kai, Feng Mingquan, Li Guoyun, Wu Yazhen, Li Furong

2017, 38(S1): 125-128. doi: 10.13832/j.jnpe.2017.S1.0125

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The Charpy impact specimen and 0.5T-CT specimen extracted from a certain domestic RPV material are irradiated in HFETR with fast neutron（E＞1 Me V）influence of 3.0×10¹⁹ cm^-2. In accordance with unirradiated and irradiated Charpy impact test, the measured ΔRT_NDT is 48℃, while, according to the fracture toughness test in ductile-brittle transition region, the measured ΔT₀ is 53℃, these actual measurement results demonstrate that the irradiation embrittlement effect is obvious. The RT_T0 based on the fracture mechanics methodology to be an alternative indexing to RTNDT could further exploit the underlying safety margin of RPV and improve the economy of NPP.

Fracture Mechanism of Neutron Irradiated B₄C/Al Composite Material

Xi Hang, ZHang Haisheng, Wu Lu, SuN Kai, Peng Yanhua, MO Huajun

2017, 38(S1): 129-132. doi: 10.13832/j.jnpe.2017.S1.0129

Abstract(17) PDF(0)

Abstract:
To get the mechanical properties of B₄C/Al material, the tensile performance and hardness of neutron irradiated B₄C/Al were studied. Fracture appearance was investigated by scanning electron microscope. The result shows that irradiated Rockwell hardness was raised, and the yield and the tensile strength increased 37 MPa and 32 MPa, respectively. The elongation at break descended 3.6%. The B₄C/Al material was brittle fracture before and after irradiation. Dislocation density increased and aluminum atoms in the enrichment of boron carbide particles surface diffusion was a major cause for the mechanics performance improvement of the boron aluminum material.

Applicability of Load Separation Method for Domestic A508-Ⅲ Steel Fracture Toughness Tests

Lei Yang, Li Guoyun, SuN Kai, ZHang Haisheng, Xi Hang

2017, 38(S1): 133-135. doi: 10.13832/j.jnpe.2017.S1.0133

Abstract(9) PDF(0)

Abstract:
In this paper, the fracture toughness of domestic A508-Ⅲ steels for 1/2T-FFCT specimens were studied by the load separation of normalization method. J-R resistance curve and J_Q value of domestic A508-Ⅲ steels were obtained, and the results were evaluated by ASTM E1820 and GB/T21143. Fracture toughness data of one irradiated A508-Ⅲ steel was studied using the normalization method, which studied the applicability of load separation method for fracture toughness tests of domestic A508-Ⅲ steels.

Study on Microscopic Structure of RPV Weld Material EG-F2N Steel after Irradiation with Impact Abnormal Data

Peng Yanhua, ZHu Wei, Tang Hongkui, Yang Fan, Feng Mingquan, Wen Bang

2017, 38(S1): 136-139. doi: 10.13832/j.jnpe.2017.S1.0136

Abstract(19) PDF(0)

Abstract:
The microstructure of post-irradiated EG-F2 N steel for reactor pressure vessel weld material after impact was observed by scanning electron microscopy（SEM） and metallographic microscope（OM）. The results show that the reason of the abnormal impact data was related to the plenty of holes on the fracture surface of samples. With the increasing of the number of holes, the effective work area was decreased and the stress concentration was appeared, thus decreasing the strength and toughness of welded specimen significantly. Additionally, a few Al₂O₃ impurities which level is considered less than 0.5 were found in some small holes. So, the main cause of low impact energy of welded specimen is not the Al₂O₃ impurities, but the big hole which is caused by improper operation of welding process.

In-situ Heating Investigation on Lattice Evolution of Zircaloy-4 during Its Oxidation Process

Wang Zhen, ZHou Bangxin, Zhu Wei, Wen Bang, Tang Hongkui, Fang Zhongqiang

2017, 38(S1): 140-144. doi: 10.13832/j.jnpe.2017.S1.0140

Abstract(12) PDF(0)

Abstract:
In-situ heating in the HRTEM using zircaloy-4 TEM thin foil specimen was carried out to investigate the oxidation behaviour of zirconium alloy at the initial stage, which provides a new view for the study of corrosion mechanism of zirconium alloys. The result shows that, the oxygen firstly is solid dissolved into the zirconium lattice to form oxygen-rich regions, and with the increasing of oxygen content, the deformation of the lattice increased and compressive strain is produced to promote the lattice modulation and the formation of the fcc structure ZrO sub-oxide. The sub-oxide is consisted of domains with similar crystal orientations and finally transformed to m-ZrO₂.

Application of Coordinate Measuring Technique in Hot Cell

Xiong Yuanyuan, Ren Liang, Jiang Linzhi

2017, 38(S1): 145-149. doi: 10.13832/j.jnpe.2017.S1.0145

Abstract(16) PDF(0)

Abstract:
The research presented in this paper focuses on the dimension measurement of nuclear fuels or materials in the hot cell using coordinate measuring machine（CMM）, which was structural adjustment to adapt to the hot cell environment. The automatic measurement was achieved by means of using coordinates to establish the position and sectional measuring with special fixture. Experiment measurements involve the thickness of plate-type fuel element（FE）, the diameter and ovality of tubular fuel element, shape dimensions of fuel assembly（FA）, and other dimensions. The research results showed that the coordinate measuring technique could be applied in the hot cell for post irradiation dimension measurement, which applied to measure different dimensions with a high measure precision of 2 μm.

Post-Irradiation Dimension Examination of Tubular Fuel Assembly

Ren Liang, Xiong Yuanyuan, Jiang Linzhi, Chen Zhe, Kuang Liuwei, Guo Chengming, Yu Feiyang, Yin Chunyan

2017, 38(S1): 150-153. doi: 10.13832/j.jnpe.2017.S1.0150

Abstract(13) PDF(0)

Abstract:
This paper mainly introduces the post-irradiation dimension measurement contents, methods and results of the new tubular fuel assembly, and makes a preliminary evaluation of the measurement results. The results show that the total length of the fuel assembly is slightly increased compared with that before irradiation, and the average growth rate is 0.71 mm. There is no change in degree of bending and twisting after irradiation. The maximum growth of margin distance is 0.17 mm and minimum growth is 0.07 mm. The water gap of each layer showes a decreasing trend, and the maximum decrease is 0.33 mm. The maximum outer diameter of the fuel pipe increases is 0.13 mm, and the maximum inner diameter of the fuel pipe decreases is 0.20 mm.

Verification of Neutron Flux Calculation Method for HFETR

Wang Hao, Xiang Yuxin, Xu Taozhong, ZHang Ping, ZHu Lei

2017, 38(S1): 154-156. doi: 10.13832/j.jnpe.2017.S1.0154

Abstract(11) PDF(0)

Abstract:
The main task for High Flux Engineering Test Reactor（HFETR） is to carry out the irradiation test of nuclear materials. The premise of the irradiation is the accurate calculation of the neutron flux. This paper introduces the neutron flux calculation method of irradiation on HFETR,and the method is verified by the test in P15 channel. The calculation data were compared with the measurement value, and the numerical results demonstrated that the deviation of calculation data was 7.14%, which satisfied the requirement of material irradiation test.

Demonstration Experiment for Graphite Material Irradiation in HFETR

Ma Liyong, Xiang Yuxin, Wang Hao, ZHang Ping, CAO Jiebao

2017, 38(S1): 157-159. doi: 10.13832/j.jnpe.2017.S1.0157

Abstract(12) PDF(0)

Abstract:
Neutron fluence irradiation goal of graphite material of solid fuel thorium molten salt reactor is 5×10²⁰cm^-2（±15%）（E＞0.1 Me V, relatively temperature 650℃±50℃）. An irradiation experiment was carried out to assure irradiation condition. The assembly of graphite material is modularized designed which is composed of the sealed part, the irradiation part, and the gas cooling system. The neutron fluence was calculated by MCNP code, and neutron indicators were installed in the graphite irradiation assembly so well. The irradiation result shows that the gas cooling system maintained the assembly temprature a little higher than the upper limit, and the MCNP method served well in neutron fluence calculating of graphite material irradiation in HFETR.

The Design and Study on Independent Temperature Compensation and High Temperature Material Irradiation Facility

Nie Liangbing, Yang Wenhua, Tong Mingyan, Xu Bin

2017, 38(S1): 160-163. doi: 10.13832/j.jnpe.2017.S1.0160

Abstract(10) PDF(0)

Abstract:
In order to realize the（650℃±50℃） material irradiation test in pile and improve the uniformity of the irradiation temperature, the design and analysis of the high temperature material irradiation facility is carried out. By selecting the appropriate material to process and manufacture the irradiation facility, the high temperature resistance performance of the irradiation facility is improved. Base on the condition experiment, the temperature independent control technique and ladder clearance temperature compensation technology are proposed to reduce the irradiation temperature difference, the critical dimension of the irradiation facility is determined, and the（650℃±50℃） irradiation temperature is realized.

Design Optimization of ~³He Gas Screen Structure in Power RAMP Irradiation Test Device

Zhang Liang, Qiu Liqing, Tong Mingyan, SuN Sheng, Yang Wenhua, Wang Hai, Li Binglin

2017, 38(S1): 164-169. doi: 10.13832/j.jnpe.2017.S1.0164

Abstract(12) PDF(0)

Abstract:
Different designs of ~³He gas screen structure are modeled and analyzed with MCNP code along with the PRT（Power Ramp Test） device installed in a HFETR core. Based on the neutronics computation results, the designs are evaluated quantitatively with a multi-objective evaluation method. The results show that when the location of the ~³He gas screen is fixed, the design with a thicker ~³He gas layer has a lower fuel rod power. Variation range of the test fuel rod power is acceptable when the ~³He gas screen has a ~³He gas layer with the thickness as 2 to 5 mm. When the distance from the fuel rod to the ~³He gas screen with a 3 mm thick gas layer is larger, the ~³He gas screen has less control of the test fuel rod power for a higher thermal neutron flux level in the rod and a larger reactivity addition during the test. Based on the marks by a quantitatively evaluation method, the optimal design of ~³He gas screen structure with a 3 mm thick ~³He gas layer and 2 mm thick annual inner flow channel, is identified for a PRT device and a typical HFETR core arrangement.

Effect of Burnup Depth on Porosity in Core of U₃Si₂-Al Dispersed Fuel Element

He Wen, Wu Xiaoyong, Wu Lu, Wen Bang, Zhu Wei, Zhang Wei, Pan Rongjian, Wang Zhen, Huang Weijie

2017, 38(S1): 170-174. doi: 10.13832/j.jnpe.2017.S1.0170

Abstract(9) PDF(0)

Abstract:
Thermal/mechanical properties of the fuels are dominantly affected by the fission pores produced from fuel particles during irradiation. In this paper, the effect of U₃Si₂ fuel particle on U₃Si₂-Al dispersed fuel were studied using the optical microscope（OM）, scanning electron microscopy（SEM） and energy dispersive spectrometer（EDS）. The microstructure of U₃Si₂ was observed. Additionally, the morphology, size and distribution of pores of U₃Si₂ were analyzed statistically. The results show that when the fission density increases from 2.34×10²⁷f/m~3 to 3.74×10²⁷f/m~3, the gas morphology in the U₃Si₂ fuel particles is globular and without great change. However, the average pore size and porosity caused by the fission pores increase with the fission density, which go through two stages: when the fission density of the fuel particles increases from 2.34×10²⁷f/m~3 to 3.19×10²⁷f/m~3, the average pore size and porosity are with steady-state growth; when the fission density of the fuel particles increases from 3.19×10²⁷3.74×10²⁷f/m~3, the average pore size and porosity increase rapidly.

Technical Research on PWR Fuel Assembly In-Pile Test

Ru Jun, Pang Hua, JiAO Yongjun, Xu Dan, Wang Kun, Liu Yanghua, Wang Haoyu

2017, 38(S1): 175-177. doi: 10.13832/j.jnpe.2017.S1.0175

Abstract(19) PDF(0)

Abstract:
PWR fuel assembly in-pile test is the most important procedure to develop a new fuel assembly. In this paper, the general requirements to draw up an irradiation plan and post irradiation examination are presented. The leading test fuel assemblies exhibit satisfactory performance, which indicates that the irradiation and examination plans are proper. The plans can be also used for reference with regard to other fuel assembly irradiation.

2017 Vol. 38, No. S1