核动力工程

Modelling Analysis of Non-uniform Flow and Heat Transfer in Parallel Rectangular Channels under Flow Blockage Condition

Chen Jiayue, Wang Zefeng, Wang Xiaoyu, Chen Huandong

2023, 44(S1): 1-8. doi: 10.13832/j.jnpe.2023.S1.0001

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Abstract:
In order to establish the non-uniform flow and heat transfer modelling method for parallel rectangular channels, and provide a new modelling method and tool for the safety analysis of plate-type fuel, one dimensional two-fluid model coupled with two-dimensional heat conduction of fuel are used to develop a thermal-hydraulic transient analysis code for non-uniform flow and heat transfer simulation under flow blockage condition. The flow distribution and fuel temperature field under different blockage conditions are obtained by numerical simulation, and the two-dimensional heat conduction effect of the fuel under four different power distributions is also studied. The simulation results show that the flow and heat transfer in parallel channels would redistribute after blockage. And the two-dimensional heat conduction model predictes a more uniform temperature profile of fuel plate cross-section. Thus, the thermal-hydraulic transient analysis code developed in this paper can be unsed to simulate the non-uniform flow and heat transfer for the plate-type fuel.

Conceptual Design of Supercritical Water-cooled Reactor CSR150

Ning Zhonghao, Wang Lianjie, Lu Di, Xia Bangyang, Huang Yanping, Chen Xing

2023, 44(S1): 9-13. doi: 10.13832/j.jnpe.2023.S1.0009

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Abstract:
Supercritical water-cooled reactor (SCWR) is regarded as one of the GEN IV nuclear reactors. Based on the conceptual design of CSR1000 proposed by NPIC, the conceptual design of supercritical water-cooled demonostration reactor CSR150 is proposed. The core design of CSR150 is studied in this paper. The core consists of 45 fuel assembles, and the fuel enrichment zoning and two-pass coolant flow scheme are adopted to increase the coolant outlet temperature and reduce the fuel cladding temperature. The study shows that the key parameters such as power distribution and fuel clad temperature meet the design objectives and criteria of CSR150.

Dynamic Simulation of Debris Bed Melting Process during the Hypothetical Severe Accident of HPR1000

Lyu Chao, Li Gen, Yan Junjie

2023, 44(S1): 14-20. doi: 10.13832/j.jnpe.2023.S1.0014

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Abstract:
At the late-phase of nuclear reactor severe accident, the melting process of debris bed in the reactor pressure vessel (RPV) lower head has significant impact on internal heat transfer characteristics, heat flux distribution on vessel wall and vessel wall ablation. In this study, based on the Ansys Fluent, the phase change model and large eddy simulation (LES) turbulence model were used to study the dynamic melting process of debris bed during the hypothetical severe accident of Hua-long pressurized reactor 1000 (HPR1000). The variations of temperature distribution, velocity field and wall ablation during the molten pool formation were predicted. The results showed that the heating rate decreased and tended to be stable after the melting of the debris bed began. The temperature distribution in the pool gradually became relatively uniform in the middle and upper parts, with a large temperature gradient at the bottom. With the increase of the decay heat, the pool part with uniform temperature expanded downward. Although the heat flux distribution on the wall inside was lower than the critical heat flux (CHF) at the corresponding outer position, wall ablation was still observed. The ablation first occurred on the inside of the wall near the surface of the debris bed and gradually spread downward. The area and depth of the ablation increased with the shortening of the debris dry-out time since reactor shutdown. The calculation results here can provide reference for the study of phase change heat transfer in the debris bed and the integrity of the RPV.

A Study on Safety Analysis of Heat Pipe Cooled Reactor Based on Unmanned Underwater Vehicle

Xu Shihao, Gou Junli, Shan Jianqiang, Ouyang Zeyu, Wang Zheng

2023, 44(S1): 21-28. doi: 10.13832/j.jnpe.2023.S1.0021

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Abstract:
Based on a compact heat pipe-cooled reactor carried by an unmanned underwater vehicle, a complete safety analysis model of heat pipe cooled reactor is established and optimized in this paper, which mainly includes core power transient model, cold start-up model of high temperature heat pipe and two-dimensional heat pipe grid model. The passive residual heat removal system under accident condition is also designed. A heat pipe-cooled reactor safety analysis program was developed based on the established model, and the program's calculated results were compared and verified with published experimental data of cold start-up and stable operation. The verification results showed good agreement between the program's calculated results and experimental data, demonstrating the accuracy of the program and the reliability of the predicted results. The typical accident of the research object was analyzed by using the program, and the highest temperature was calculated to be 1085K under the heat sink loss accident condition, with a delay of reactor shutdown for 3s and a delay in putting residual heat removal system into operation for 6s, and it's below the maximum operating temperature of the heat pipe. The transient response of the reactor temperature with a step-in positive reactivity insertion of 0.47$ and a linear reactivity insertion of ±0.05$ was also calculated, and the highest temperature was below the maximum operating temperature of the heat pipe. Under feedback regulation, the reactor reached a new steady state at a higher power level, demonstrating the good inherent safety of the reactor design scheme.

Experiment Research on Integral Hydraulic Simulation of ACP100 Reactor

Ding Lei, Chen Xing, Wang Dianle, Xu Jianjun, Sui Xi, Fang Ying, Meng Yang

2023, 44(S1): 29-34. doi: 10.13832/j.jnpe.2023.S1.0029

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Abstract:
Advanced Small Modular Pressurized Reactor (ACP100) is a new integrated small-scale reactor. Its once-through steam generators and reactor coolant pumps are directly integrated into the pressure vessel, and the compact design leads to a complex internal flow field. A 1/3 scaled-down model was used to simulate the entire internal flow field of the ACP100 reactor and carry out integral hydraulic simulation cold experiment for the reactor. In the experiment, the total pressure drop and segmented pressure drop of the model were obtained, as well as the total resistance coefficient and the segmented resistance coefficients of the main flow paths. The flow distribution factors of each fuel assembly at the core inlet were also acquired. The experiment results showed that the flow inside the main channels had entered the second self-modeling zone, and the flow pattern, velocity distribution and resistance coefficient of the fluid were consistent with those of the prototype reactor. After entering the second self-modeling zone, the total resistance coefficient of the model remained constant at 8.02, which can be used to calculate the pressure drop of the prototype directly. Under rated operating condition, all the distribution factors at the core inlet ranged from 0.91 to 1.08, meeting the design requirements. Besides, the flow distribution of LOFA simulation was uniform, indicating that the flow distributor had good rectifying effects.

Investigation on Flow Field Characteristics of Rod Bundle Channel under Rolling Condition

Qi Chao, Li Xin, Tan Sichao, Cheng Kun, Qiao Shouxu

2023, 44(S1): 35-39. doi: 10.13832/j.jnpe.2023.S1.0035

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Abstract:
The reactor is in an unsteady state under ocean conditions, which will cause tilt, swing, undulation and other motions of the reactor. These motions will introduce additional inertial force fields in the rod bundle channel, bringing additional influence on the flow field in the rod bundle channel. Therefore, it is necessary to study the rod bundle channel under rolling conditions. In this paper, based on the particle image velocimetry (PIV) technology, the research on the flow field distribution characteristics of the rod bundle channel with a pitch-to-diameter ratio of 1.326 under rolling condition is carried out. The difference of flow field distribution between steady state and transient state under the same flow rate is compared, and the flow field distribution characteristics at different positions in the rod bundle channel under the same acceleration are analyzed. The experimental results show that the rolling motion has a small effect on the middle of the rod bundle channel, and has a greater impact on both sides of the channel. The velocity field on both sides of the channel fluctuates periodically, with an anti-phase wave. In the case of low flow, reverse flow phenomenon will occur, but the spacer grid has no effect on the lateral speed upstream. Studies have shown that the flow field change caused by rolling motion is quite different from that caused by pulsating flow, in which the velocity field change caused by pulsating flow is uniform and pulsating, while the velocity field caused by rolling presents anti-phase fluctuation on both sides of the channel.

Influence of Different Turbulence Models on Simulation of Lead-Bismuth Solidification

Zeng Chen, Zhang Rui, Liu Maolong, Zhang Weihao, Li Junlong, Liu Limin, Liu Li, Gu Hanyang

2023, 44(S1): 40-45. doi: 10.13832/j.jnpe.2023.S1.0040

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Abstract:
To study the influence of different turbulence models and turbulence Prandtl number (Pr_t) models on the simulation of lead-bismuth solidification behavior, this paper uses FLUENT to simulate the flow solidification behavior of lead-bismuth (Pb-Bi) inside a tube. The simulation results show that although the differences between the shear stress transfer (k-ω SST), k-ε and Reynolds stress (RSM) models in simulating Pb-Bi heat transfer can be ignored, there are significant differences in the simulation of the temperature and pressure fields during the phase change, thus the turbulence model should be carefully selected. In addition, the study on the simulation of Pb-Bi solidification using different Pr_t models shows no significant difference.

Research on Reproduction Method of CRUD Depositions Based on Anodic Oxidation

Liu Yan, Liu Xiaojing, Du Sijia, Wang Jiageng, He Hui

2023, 44(S1): 46-50. doi: 10.13832/j.jnpe.2023.S1.0046

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Abstract:
In order to establish the time and spatial distribution model of boron and provide support for the experiment of boron transport and heat transfer within the Chalk River Unidentified Deposit (CRUD) depositions, the reproduction method of CRUD depositions was studied based on anodic oxidation method. With reference to the morphology data of the real CRUD depositions in the PWRs, the effective control of the pore size can be realized by adjusting the ammonium fluoride concentration, water concentration, oxidation voltage and oxidation time of the electrochemical system. Characterization analysis shows that the reproduced CRUD deposition samples can fully cover the range of morphological parameters of the real CRUD depositions. Therefore, the reproduction method established in this study can be used for the reproduction of CRUD deposition samples.

Kinetic Mechanism Study of Lead-Bismuth Alloy Solidification in Lead-Water Interaction

Zhang Lin, Liu Dalin, Deng Chang, Liu Xiaojing

2023, 44(S1): 51-56. doi: 10.13832/j.jnpe.2023.S1.0051

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Abstract:
To obtain the kinetic mechanism of lead-bismuth alloy solidification during the lead-water interaction and understand the microscopic dendrite growth process, the average flow velocity equation of lead-bismuth alloy between dendrites was developed by analyzing the natural convection of lead-bismuth melt and its influence on the solidification front. In addition, the dendrite growth process was simulated by using the phase field method. The results showed that the fast growth zone of dendrites was markedly tilted toward the incoming flow direction under the joint action of flow fields in two directions. This work can provide kinetic mechanism analysis for lead-bismuth alloy solidification, and provide the theoretical basis for the safe operation of lead-based reactors.

Numerical Investigation of LBE Flow and Heat Transfer Characteristics in Helical-coiled Tube Bundles with Different Coil Strategies

Shen Cong, Liu Maolong, Cheng Kun, Liu Limin, Xu Ziyi, Gu Hanyang

2023, 44(S1): 57-61. doi: 10.13832/j.jnpe.2023.S1.0057

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Abstract:
Helical tube once-through steam generators (H-OTSG) are widely used in the design of liquid metal reactors, in which adjacent radial tube bundles can be coiled in the same direction or opposite direction, and different coil strategies will affect the flow behavior on the shell side of the steam generator. To explore the lead-bismuth eutectic (LBE) flow and heat transfer characteristics in helical tube bundles with different coil strategies, the shear stress transport (SST k-ω) model, turbulence model and Kays turbulent Prandtl number (Pr_t) model are used for numerical simulation. First, the numerical method is validated by existing experiments of liquid metal flow cross tube banks. Then, helical-coiled tube bundles with the same coiling direction and alternate coiling direction are established, and the differences in heat transfer and flow resistance are compared. Finally, the reason for the differences is analyzed from the perspective of the flow field. The results show that the flow resistance and heat transfer in the helical-coiled tube bundle with alternate coiling direction are 7.1% and 4.4% higher than those with the same coiling direction respectively. This is due to the stronger turbulent mixing and more uniform velocity field in the alternate-coiled bundle.

Study on Stability of Natural Circulation of LBE in Rectangular Circuit

Wang Xin, Kuang Bo, Wang Shuting, Hu Wenjun, Ren Lixia

2023, 44(S1): 62-68. doi: 10.13832/j.jnpe.2023.S1.0062

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Abstract:
In order to analyze the stability of the natural circulation flow of lead-bismuth eutectic (LBE) in the rectangular circuit, and to obtain the factors affecting the natural circulation flow and the law of action, a one-dimensional system program FRTAC is developed by time-domain method to numerically calculate the lead-bismuth natural circulation, and its stability characteristics were determined by analyzing and processing the dynamic time series. The stability domain and stability boundary of the lead-bismuth natural circulation under the relevant operating conditions, the flow stability characteristics of the lead-bismuth natural circulation loop, as well as the influencing factors and laws are obtained. The results show that increasing the length of the heating section of the loop will increase the stability of the system, while increasing the loop diameter and the height difference between the hot and cold cores will decrease the stability margin and weaken the stability of the system.

Neutronic and Thermal-Hydraulic Performance Analysis of Helical Cruciform Fuel Rods

Zhang Tao, Han Wenbin, Shen Pengfei, Huang Shanfang, Wang Kan

2023, 44(S1): 69-74. doi: 10.13832/j.jnpe.2023.S1.0069

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Abstract:
To analyze the neutronic and thermal-hydraulic performance of helical cruciform fuel (HCF) rods, numerical simulations using the CAD (computer-aided design)-based Reactor Monte Carlo code RMC and the commercial computational fluid dynamics (CFD) software Fluent were conducted, and the results were compared with those of traditional cylindrical and untwisted cruciform fuel rods. The results show that the helical cruciform structure slightly reduces the reactivity and increases the radial power peaking factor. Compared with cylindrical fuel rods, the HCF rods can enhance coolant mixing and heat transfer due to their transverse flow characteristics. In the 7-rods assembly calculation, the mean and peak temperatures of HCF rods are reduced by about 4 K.

Analysis of Potential Impact of ATFs on Reactor Safety under Shaft-Stuck Accident

Wu Hexin, Jin Desheng, Gou Junli, Shan Jianqiang, Cheng Yi

2023, 44(S1): 75-80. doi: 10.13832/j.jnpe.2023.S1.0075

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Abstract:
To analyze the potential impact of accident tolerant fuel (ATF) on reactor safety under pump shaft-stuck accident, China's improved three-loop pressurized water reactor CPR1000 was used as reference power station to carry out second development based on the system analysis code NUSOL-SYS. The performance of CPR1000 with different ATF combinations under pump shaft-stuck accident was studied, and the sensitivity analysis was carried out to study the change of heat transfer coefficient and critical heat flux (CHF) caused by the change of ATF cladding surface characteristics. The results show that the change of heat transfer coefficient and CHF caused by the change of ATF cladding surface characteristics has a great influence on the maximum temperature of pellets and the peak cladding temperature (PCT). ATF pellets with high thermal conductivity can greatly reduce the pellet temperature, while ATF materials with high specific heat capacity can reduce the PCT.

Design and Performance Investigation of Supercritical Carbon Dioxide Ejector

Feng Mengjiao, Liu Minyun, Huang Shanfang, Huang YanPing

2023, 44(S1): 81-87. doi: 10.13832/j.jnpe.2023.S1.0081

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Abstract:
In order to minimize the negative impact of leakage gas produced on the stability of the supercritical carbon dioxide cycle without additional mechanical work consumption, an ejector that can pressurize and transport the leakage gas back to the cycle was designed in this paper based on the constant pressure mixing theory and the double choked critical state. The computational fluid dynamics software Fluent was used to simulate the model numerically, analyze the performance of the ejector, and investigate the effect of size parameters on the performance. The results show that the four-stage ejector can sequentially pressurize the 0.5 MPa leakage gas to 2.0, 4.4, 6.0 and 8.0 MPa, realizing the recovery and utilization of the leaked gas. The entrainment ratio is not affected by the back pressure when the back pressure is less than the critical pressure. The entrainment ratio decreases sharply with the increase of the back pressure when the back pressure is greater than the critical pressure. The entrainment ratio increases with the diminution of the inlet area. The entrainment ratio increases first and then decreases with the increase of the contraction angle of the nozzle, and reaches maximum while the contraction angle is 20°.

Numerical Research on Anti-corrosion Properties of Rod Bundle Channel

Wang Suhao, Li Ying, Yue Nina, Guo Liang, Xiao Hui, Lou Ruifan, Zhuo Wenbin

2023, 44(S1): 88-94. doi: 10.13832/j.jnpe.2023.S1.0088

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Abstract:
In order to obtain the formation of anti-corrosion layer on the surface of fuel assembly channels in advanced reactors and provide support for the analysis of reactor operation strategy, this paper puts forward an oxygen transport calculation model, and combined with Computational Fluid Dynamics method, the anti-corrosion layer generated in typical 19-rod bundle channels of fuel assemblies is analyzed. The flow field and temperature field in the rod bundle channel, the oxygen concentration distribution of the rod bundle under two kinds of inlet oxygen concentrations and three kinds of operation time, and the formation of the anti-corrosion layer on the rod bundle surface are obtained. The results show that the anti-corrosion layer in the bundle channel is mainly related to the temperature, initial oxygen concentration and operation time. For the existing model, the vicinity of the contact point between spacer and rod is the main area where the anti-corrosion layer is difficult to form, which needs to be paid attention to. The calculation method and results in this paper will provide support for the evaluation of reactor operation strategy.

Study on Reactivity Control of CSR150

Lu Di, Wang Lianjie, Xia Bangyang, Huang Yanping, Yao Lei, Liu Xinyao, Zhou Yajing

2023, 44(S1): 95-100. doi: 10.13832/j.jnpe.2023.S1.0095

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Abstract:
The supercritical water-cooled demonostration reactor CSR150 uses the reactivity control technology in CSR1000 for reference: reactivity control is realized through burnable poisons and control rods. Based on the selection of erbium oxide as burnable poison, this paper puts forward a design method of ¹⁶⁷Er enrichment to reduce the reactivity penalty caused by erbium oxide at EOL. The control rod design of CSR150 is studied, and the zoning design of the control rod is proposed. Boron enrichment is used as the safety rod material to improve the reactivity control ability. The evaluation results of key parameters of the core show that the proposed reactivity control scheme meets the design requirements of CSR150.

Thermal-mechanic Analysis on COOL Blanket for CFETR

Jiang Kecheng, Yu Yi, Ma Xuebin, Chen Lei, Liu Songlin

2023, 44(S1): 101-107. doi: 10.13832/j.jnpe.2023.S1.0101

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The supercritical carbon dioxide (sCO₂) liquid lead-lithium blanket (COOL) is a candidate for China Fusion Engineering Test Reactor (CFETR), and its main functions are tritium breeding, neutron radiation shielding and energy conversion to generate electricity. COOL cladding needs to bear loads such as coolant pressure, thermal stress, gravity and electromagnetic force under normal operating conditions. In this paper, the thermal and mechanical performance of the equatorial outboard blanket module in the COOL blanket segment are analyzed by ANSYS finite element method. The results show that the maximum temperature of various materials of the blanket under normal operating condition does not exceed the upper limit, and the structural stress could meet the ITER SDC-IC design standard. The analysis results can provide important reference and data support for the following iterative optimal design of the blanket.

Prediction of Fluid Critical Point Based on Molecular Dynamics Simulation

Zhao Xuebin, Huang Yanping, Ye Lyu

2023, 44(S1): 108-112. doi: 10.13832/j.jnpe.2023.S1.0108

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The critical points of CO₂ and H₂O were predicted using molecular dynamics simulations, and relatively accurate critical points were obtained by extrapolating properties under the vapour-liquid equilibrium. For CO₂, the simulations were carried out with TraPPE and a coarse-grained model, SAFT. The results simulated by TraPPE are in better agreement with experimental data of National Institute of Standards and Technology (NIST). Potential models of SPC/E and TIP4P/2005 were used to evaluate the critical point parameters for H₂O. The results suggest that the predicted values with TIP4P/2005 are closest to the experimental data of NIST, and there are still challenges to accurately predict the saturated vapor pressure of molecular water system.

Research Progress on High Temperature Oxidation Behavior of Zirconium Cladding under LOCA Condition

Zhao Wanqian, Jia Yuzhen, Pei Jingyuan, Li Guoqing, Lyu Junnan, Zhang Junsong, Liao Jingjing, Peng Qian

2023, 44(S1): 119-124. doi: 10.13832/j.jnpe.2023.S1.0119

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Zirconium alloy is a cladding material widely used in water-cooled power reactor. The high temperature behavior of zirconium alloy cladding under the extreme accident condition of loss of coolant accident (LOCA) has become a hot topic of research and being discussed at home and abroad. This paper summarizes the worldwide current research progress on high temperature oxidation behavior of zirconium alloy. The oxidation kinetics characteristics, breakaway oxidation behavior, and the oxidation transition mechanism are described in detail. Meanwhile, the research work of Nuclear Power Institute of China for nearly recent ten years has also been overviewed. The research progress reported in this paper, especially the discussion on the oxidation transition mechanism, will provide a theoretical guidance for developing domestic new zirconium alloys in the further.

Research on Influence of Structural Parameters on Shielding Efficiency of Al-based Foam Metal

Wu Songling, Ye Zhutao, Li Aihua, Li Gang, Liu Xiaozhen

2023, 44(S1): 125-130. doi: 10.13832/j.jnpe.2023.S1.0125

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Metal foam is characterized with excellent γ-ray shielding ability and lightweight. However, the relationship between its structural parameters and the radiation shielding performance is not clear, which blocks the further optimization of shielding properties. In this paper, two simulation models of densest packing spheres are constructed, then the shielding performance of ideal Al-based foam metal with various structural parameters is calculated by Monte Carlo. It is found that the Al-based foam metal’s shielding capacity against soft γ-ray with energy below 0.24 MeV is better than that of A356. Packing density of hollow spheres is the dominant controllable route to optimize the shielding performance of Al-based foam metal: and the lighter the material is, the better the performance is within its suitable radiation shielding energy range. The mode of packing is the most critical factor to affect the shielding capacity against γ-ray from ¹³⁷Cs and ⁶⁰Co sources. Preparation of face centered cubic densest packing sphere Al-based foam metal will weaken its inferiority in shielding harder γ-ray.

Study on Anisotropy of Mechanical Properties of Zr-Sn-Nb Alloy Strip

Cui Yiran, Yang Zhongbo, Liu Ranchao, Deng Chuandong, Wang Xiaomin, Qiu Jun, Xu Chunrong

2023, 44(S1): 131-136. doi: 10.13832/j.jnpe.2023.S1.0131

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In order to explore the influence of the texture of the Zr-Sn-Nb alloy strip on its mechanical properties in different directions, the texture and mechanical properties in different directions of the finished annealed Zr-0.85Sn-1Nb-0.3Fe alloy strip were systematically analyzed by Electron Backscatter Diffraction (EBSD) and tensile test. The results showed that the typical bimodal basal texture was formed in the alloy strip, leading to the anisotropy of mechanical properties. With the increase of the angle between the loading direction and the RD (Rolling Direction), the Schmid factor of the main slip system in the strip showed a decreasing trend, which increased the opening difficulty of the slip system. Also the yield strength and yield ratio increased, while the tensile strength and work-hardening index decreased. As a result, the Zr-Sn-Nb alloy strip has relatively better stamping formability in the RD.

Corrosion Behavior of High Strength AlCrFeNi Multi-principal- component Alloy in Lead-bismuth Alloy

Huang Yunhao, Wang Jianbin, Wang Zhijun, Zhao Ke

2023, 44(S1): 137-142. doi: 10.13832/j.jnpe.2023.S1.0137

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Traditional structure materials limit the higher performance of lead-bismuth nuclear system. In order to provide high-performance structural materials for lead-bismuth reactors, a study on static lead-bismuth eutectic compatability at high temperature for the high strength Al₁₇Cr₁₀Fe₃₇Ni₃₆ multi-principal-component alloy was carried out. The results showed that the alloy formed a dense Fe-Cr-Al-O oxide film and a loose iron oxide double-layer oxide film structure in the lead-bismuth saturated oxygen environment at 500~600℃. The double oxidation film is only 1.5μm, which shows that the growth velocity of double oxidation film is slowly. The study show the Fe–Cr–Al oxide film own excellent compactness and structure stability, which prevent the lead-bismuth corrode the matrix. This study shows that the advance of AlCrFeNi multi-principal-component alloys in the lead-bismuth system, compared with the traditional ferrite/martensite stainless steel and austenite stainless steel.

Effect of Thermal Aging on Impact Toughness of 20Cr25NiNb Stainless Steel

Shu Ming, Sun Yongduo, Zheng Yuqi, Zhou Qin, Liu Xiao, Xiao Jun, Chen Luyao

2023, 44(S1): 143-146. doi: 10.13832/j.jnpe.2023.S1.0143

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In view of the degradation of alloy mechanical properties under prolonged high temperature exposure, the thermal aging study of candidate cladding materials for supercritical gas-cooled reactor was carried out. The prototype 20Cr25NiNb stainless steel for advanced gas-cooled reactor (AGR) reactor and improved alloys with different elements were subjected to thermal aging test at 650°C for 3000 h. The results of microstructure and properties showed that the impact absorbed energy (KV₂) of all alloys decreased as the thermal aging proceeded. This plasticity decrease was closely related to the evolution of second phases at high temperature. The precipitation of M₂₃C₆ and G phases along grain boundaries led to a decrease and then a slow increase in the impact toughness of the prototype alloy. The addition of W and Mo elements brought the precipitation of Laves and σ phases at grain boundaries, causing a faster decrease in KV₂; B element refined σ phase and made the impact toughness decrease less than that of the alloy without B. After adding Al, a large number of Laves and NiAl phases precipitated in the matrix, while the σ phase at grain boundaries coarsened rapidly, leading to severe material embrittlement.

Effect of Thermal Aging on Mechanical Properties of Silicon-containing Ferritic/Martensitic Steel

Liu Xiao, Wang Hui, Xiao Jun, Sun Yongduo, Liu Shuaiyang, Zhang Jinyu

2023, 44(S1): 147-151. doi: 10.13832/j.jnpe.2023.S1.0147

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Four kinds of 9Cr- ferritic/martensitic (F/M) steels with different Si contents were heat aged at 550°C (the longest time was 5000 h), and their mechanical properties such as yield strength (R_p0.2), tensile strength (R_m) and elongation (A) were tested. The relationship between microstructure and mechanical properties was studied by means of scanning electron microscope/energy dispersive spectrometer (SEM/EDS) and transmission electron microscope (TEM). The results show that the strength of 9Cr-F/M steel can be improved by adding a small amount of Si, and when the Si content (mass fraction) is 0.7%, R_p0.2 and R_m reach the maximum, but the addition of Si will promote Laves phase precipitation. Aging time (t) has a significant effect on the plasticity of 9Cr-F/M steel. When t < 2500 h, the plasticity of 9Cr-F/M steel has little change, but when the Si content is increased to 1.0%, the plasticity decreases greatly after aging for 5000 h, which is attributed to the precipitation and growth of Laves phase at the grain boundary.

Analysis of Crevice Corrosion behavior of Titanium Alloy in Boron and Lithium Media

Zhao Yuxiang, Xu Qi, Xiong Ru, Guo Xianglong, Liu Xiao

2023, 44(S1): 152-157. doi: 10.13832/j.jnpe.2023.S1.0152

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As the main structural material of new steam generator (SG), the crevice corrosion behavior of titanium alloy has been concerned. However, the crevice corrosion resistance of titanium alloy in boron and lithium media is less studied. In this paper, the crevice corrosion behavior of titanium alloys TA16 and TA17 in boron and lithium media for 5000 h was studied by means of micro corrosion loop. The crevice corrosion sensitivity of the two materials was obtained, and the composition and structure of the oxide film of titanium alloys were analyzed. The results show that no crevice corrosion is observed on the sample, which indicates that TA16 and TA17 are insensitive to crevice corrosion in boron and lithium media. There are some differences between the oxides inside and outside the crevice of titanium alloys, and the granular microcrystal FeTiO₃ outside the crevice is independent of crevice corrosion of titanium alloy.

Research on U-Zr-based Metallic Fuel Additives and Performance Improvement for Fuel-Cladding Chemical Interaction and Phase Optimization

Zhuo Weiqian

2023, 44(S1): 158-162. doi: 10.13832/j.jnpe.2023.S1.0158

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The current work focuses on the fuel additives Sb, Mo, Nb, and Ti for U-Zr-based metallic fuel in order to mitigate fuel-cladding chemical interaction (FCCI) and optimize the fuel phase. A series of U-Zr-based fuel samples were evaluated out-of-pile by diffusion couple tests and annealing tests, and characterized by scanning electron microscope (SEM), X-ray diffraction (XRD) and differential scanning calorimetry (DSC). The additive Sb was used to mitigate FCCI under high burnup. The results showed that Sb formed precipitates with lanthanide Ce, and the precipitates did not react with cladding materials. This indicates that the additive Sb is a promising candidate to solve the FCCI problem under high burnup. The additives Mo, Nb and Ti were used to optimize the γ phase transition temperature of metallic fuel. It was found that Mo and Nb could reduce the transition temperature of γ phase, and the effect of Mo was better than that of Nb. The results in this work can provide fundamental data for the advanced fuel design in the future.

Effect of Final Annealing Temperature on Microstructure and Properties of N36 Alloy Tube

Jia Yuzhen, Qiu Jun, Cheng Zhuqing, Yang Zhongbo

2023, 44(S1): 163-167. doi: 10.13832/j.jnpe.2023.S1.0163

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Abstract:
In order to optimize and control the microstructure and properties of N36 alloy tube, the effect of final annealing temperature on the microstructure and properties of N36 alloy tube was studied by analyzing the performance data of N36 alloy tube at different final annealing temperatures (520~560°C). The results show that different final annealing temperatures mainly affect the recrystallization fraction and grain size of the N36 alloy tubes, while have little influence on the second phase particles. The higher the final annealing temperature, the higher the recrystallization fraction and the larger the grain size of N36 alloy tube. With the increase of final annealing temperature, the room temperature and high temperature axial and circumferential tensile strength decreased and the elongation increased significantly, which was mainly caused by the influence of final annealing process on the recrystallization fraction and grain size of N36 alloy tube. With the increase of final annealing temperature, the corrosion resistance of N36 alloy tube is improved, and the corrosion resistance of N36 alloy tube with final annealing temperature of 560°C is obviously better than other tubes, mainly due to the highest recrystallization fraction of N36 alloy pipe with final annealing temperature of 560°C.

Study on Oxidation Behavior During Steam Oxidation of Cr-coated Zirconium Alloy at High-Temperature

Yang Hongyan, Lyu Junnan, Zhang Ruiqian, Wei Tianguo

2023, 44(S1): 168-175. doi: 10.13832/j.jnpe.2023.S1.0168

Abstract(436) HTML (35) PDF(38)

Abstract:
For systematic review of the high-temperature oxidation resistance of Cr-coated Zirconium alloy fabricated by multi-arc ion plating, the oxidation behavior during the steam oxidation of Cr coating at 800-1400°C was studied. The oxidation weight gains were obtained by simultaneous thermal analyzer. The X-ray diffractometer was used to analyze the phase structure, the scanning electron microscope was used to analyze the surface morphology and thickness of each film in cross section, and the energy dispersive spectrometer was used to analyze the element distribution. The dense Cr₂O₃ scales were formed on Cr coating after the steam oxidation at 1200℃, and the good interfacial bonding quality were observed between coating and substrate, as well as the residual Cr coating on Zr substrate were remained, and the residual Cr coating on Zr substrate were remained. The Cr coating could protect substrate for a short time at 1300℃ steam, but quickly lost its protection at 1400℃ steam. The Cr-coated Zirconium alloy has excellent high temperature oxidation resistance, and it still has good protection on substrate after the steam oxidation at 1200℃.

Study on Long-Term High Temperature Creep Properties of 12Cr-1.5W-0.6Si Alloy Pipe

He Kun, Pan Qianfu, Li Gang, Liang Bo

2023, 44(S1): 176-180. doi: 10.13832/j.jnpe.2023.S1.0176

Abstract(279) HTML (28) PDF(11)

Abstract:
To obtain the creep properties of Ferritic/Martensitic alloy, creep tests for 12Cr-1.5W-0.6Si pipes in the air environment at 450°C, 500°C and 550°C were carried out to obtain the creep time-strain curve and steady-state creep rate by using the creeptesting device. The results show that the stress index is relatively high, and the true stress index is obtained by introducing threshold stress. The creep mechanism is dislocation climbing mechanism. After creep tests at 550°C and 160MPa for 3145 h, the precipitated phase is still distributed along the grain boundary, but the lath grain widens and the precipitated phase grain coarsens, and the effect of long-term creep on the microstructure is more significant.

Effect of Composition Optimization of Ni-base Filler Alloy on Microstructure of Vacuum Brazing Joint of SiC_f/SiC Composite Ceramic

Shi Haojiang, Zhang Ruiqian, Yan Jiazhen, Li Ming, Liu Zihao, Bai Dong, He Yong, Lyu Junnan

2023, 44(S1): 181-187. doi: 10.13832/j.jnpe.2023.S1.0181

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Abstract:
In order to explore the difference of brazing process between SiC_f/SiC composites and SiC ceramics, the vacuum brazing connection between SiC_f/SiC composites and SiC ceramics was realized by using Ni-based alloy brazing filler metal to hold heat for 10 min at 1400°C in vacuum, and the composition of Ni-based filler metal was optimized according to the microstructure characteristics of weld joints. Finally, the SiC_f/SiC composite ceramic vacuum brazed joint with good weld quality and less SiC fiber damage was obtained. The results show that the reaction mechanism of SiC ceramics and SiC_f/SiC composites is the same as that of filler metal, and there is no difference in the reaction products. Due to the structural characteristics of SiC_f/SiC composites, the filler metal will be lost along the gap under the action of capillarity and react with SiC fibers to cause fiber damage during brazing. Therefore, the reaction tendency between filler metal and SiC_f/SiC composites should be reduced to avoid the quality damage of SiC fibers. The results can provide reference for the composition of filler metal and brazing process design of SiC_f/SiC composites.

Study on Low-cycle Fatigue Test Method for Zr-Sn-Nb Alloy Welded Sheets

Qi Yanqiang, Li Shunping, Peng Qian, Yan Meng, Chen Le, Dai Xun

2023, 44(S1): 188-193. doi: 10.13832/j.jnpe.2023.S1.0188

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Abstract:
In order to study the low cycle fatigue (LCF) properties of Zr-Sn-Nb alloy welded sheets, microstructure simulation and finite element method were combined based on the LCF test method of Zr-Sn-Nb alloy sheets in this study. The microstructure, microhardness and uniaxial tensile properties were compared and evaluated, and it turned out that it was basically feasible to simplify the complex microstructure of welded joint funnel sample by microstructure simulation. Finally, the stress-strain transformation of welded sheet funnel sample was established based on coupling, and the LCF properties of welded sheets were evaluated by this method. The results show that the LCF test method established in this study for Zr-Sn-Nb alloy welded sheets can be used to evaluate the LCF properties of the welded sheets.

Time-Limited Aging Analysis for Operating License Extension of HFETR

Deng Yunli, Liu Peng, Li Songfa, Wan Qinfang, Dai Yubing, Kang Changhu, Li Jiwa

2023, 44(S1): 194-198. doi: 10.13832/j.jnpe.2023.S1.0194

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Abstract:
In order to complete the safety demonstration of High Flux Engineering Test Reactor (HFETR) for the Operation License Extension (OLE) application, it is necessary to provide a complete list of time-limited aging analysis (TLAA), and demonstrate by calculation or analysis that each TLAA meets the requirements of safe operation in the extended operation period. Based on the TLAA experience of pressurized water nuclear power plants in the United States, this paper selects general TLAAs and potential TLAAs by means of “material-environment-influencing factors”, and identifies HFETR-specific TLAAs. Therefore, 9 TLAAs need to be supplemented, recalculated or analyzed for HFETR. The results of each TLAA evaluation show that the security of HFETR can be guaranteed during 10-year license extension period.

Study on Method of Aging Management During Operation for High Flux Engineering Test Reactor

Zhao Peng, Pan Ruian, Liu Zhen, Han Liangwen, Gao Yedong, Lai lisi, Sun Biao

2023, 44(S1): 199-204. doi: 10.13832/j.jnpe.2023.S1.0199

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Abstract:
In order to effectively carry out aging management of High Flux Engineering Test Reactor (HFETR), it is necessary to implement aging management activities in accordance with effective aging management methods. Through research on aging management scope screening, aging recognition, and aging management activities implementation methods and strategies, and according to the aging management practice and experience of operation license extension (OLE) application for HFETR, a systematic aging management system based on Deming cycle "plan-implementation-inspection-action (PDCA)" has been established for the reactor operation stage. The aging management review of HFETR in the first operation license extension application shows that the aging management method of HFETR is effective and can provide reference for the aging management of research reactors.

Technical Research and Practice of Aging Management Review for Research Reactor

Li Songfa, Zhao Guang, Lyu Yunhe, Li Wenyu, Li Yun, Yan Xiongwei, Yao Liang

2023, 44(S1): 205-210. doi: 10.13832/j.jnpe.2023.S1.0205

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Abstract:
In the Operation License Extension (OLE) application of High Flux Engineering Test Reactor (HFETR), in order to demonstrate the effectiveness of reactor aging management, the aging management review (AMR) technology for research reactor was studied by using the strategy of analysis principle based on aging management guides of International Atomic Energy Agency (IAEA) and implementation process based on the license renewal (LR) process of nuclear power plants in the United States. The safety demonstration basis, the method of scoping and screening, and the process of AMR were determined. The AMR process demonstrated that the identified aging effects could be effectively managed by the HFETR Aging Management Programs (AMPs), ensuring an acceptable level of safety during extended operation life of the reactor. The AMR method provided in this paper can provide an engineering application reference for the implementation of OLE application activities and the establishment of aging management system for domestic research reactors.

Research and Application of Operation License Extension Technical Route for Research Reactor

Liu Peng, Li Songfa, Chen Qibing, Han Shanbiao, Deng Yunli, Lai Lisi

2023, 44(S1): 211-216. doi: 10.13832/j.jnpe.2023.S1.0211

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Abstract:
In order to support the Operation License Extension (OLE) application for High Flux Engineering Test Reactor (HFETR), a set of OLE application methods suitable for HFETR was established by analyzing the license renewal (LR) method of the United States and the long-term operation (LTO) method of the International Atomic Energy Agency (IAEA), with adoption of the technical route of "based on IAEA in principle and learning from the United States in operation". This method was used to demonstrate the safety of HFETR with Periodic Safety Review (PSR), Aging Management Review (AMR), and Time-Limited Aging Analysis (TLAA). The results showed that the method applied in the OLE application of HFETR could meet the requirements of nuclear safety supervision for research reactors in China, and the application finally passed the review of the regulatory authority and obtained the extended operation license.

Determination of Aging Management Scope for HFETR

Cai Wenchao, Li Songfa, Deng Yunli, Li Changxiang, Feng Haozhi, Qin Fujun, Lu Xing

2023, 44(S1): 217-222. doi: 10.13832/j.jnpe.2023.S1.0217

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Abstract:
In order to systematically screen out the aging management review (AMR) objects required by the application for Operating License Extension (OLE) of High Flux Engineering Test Reactor (HFETR) and determine the aging management scope of HFETR, the assessment scope delineation and object screening principles of the nuclear power plant AMR of the International Atomic Energy Agency (IAEA) and the US Nuclear Regulatory Commission (NRC) were analyzed, and the screening principles of HFETR aging management items based on the requirements of research reactor aging management was established. In the process of operation, drawing on the application experience of the US Nuclear Power License Renewal (LR), specific screening processes were established for mechanical components, electrical instrument components and structures. In the implementation process, the AMR objects required by OLE application were systematically screened out by grouping items, and the aging management scope of HFETR was determined. The method and application process described in this paper can provide guidance for the determination of aging management scope of research reactors in China.

2023 Vol. 44, No. S1